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Shanghai, Nuclear fuel cycle Asian Symposium

Shanghai, Nuclear fuel cycle Asian Symposium
Reprocessing of Spent Nuclear Fuel
Radionuclides, Technetium, Actinides, Transuranium elements

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2013 abstract book ASNFC-Shanghai 2013 abstract book ASNFC-Shanghai Document Transcript

  • Contents Plenary Lectures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 Some Hot Issues on Nuclear Energy Chemistry in China . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 . Present Status of Nuclear in Japan after the Accident of Fukushima Daiichi . . . . . . . . . . . . . . 3 Session 1: General Issues on Nuclear Energy and Fuel Cycle . . . . . . . . . . . . . . . . . 5 . 1.1 Envision of World Nuclear Energy / Fuel Cycle Development and China's Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 . 1.2 The Role of Advanced Reprocessing Technology on 3S (Safety, Security, and Safeguards) in Nuclear Fuel Cycle and Radioactive Waste management . . . . . . . . . . . . 7 . 1.3 Flexible Fuel Cycle Initiative to Cope with the Uncertainties after Fukushima Daiichi NPP Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .8 1.4 Advanced Fuel Cycle : Status and Technology Development at KAERI . . . . . . . . . . . . . . 9 . 1.5 The Nuclear Education and Training Program at University of California Irvine . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .10 Session 2: Basic Chemistry of Actinides and Fission Products . . . . . . . . . . . . . . .11 2.1 Utilization of Technetium and Actinide Compound Synthesis and Coordination Chemistry for the Nuclear Fuel Cycle: Exploring Separations, Fuels, and Waste Forms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 . 2.2 Heptavalent State of Transuranium Elements, Technetium and the Other Elements of the Periodic Table . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.3 DFT Study on a Trivalent Uranium Complex Promoted Functionalization of Carbon Dioxide and Carbon Disulfide . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 . 2.4 Using Phosphonates to Probe Structural Differences Between the Transuranium Elements and Their Proposed Surrogates . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 . 2.5 From Thorium to Curium: Unprecedented Structures and Properties in Actinide Borates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .18 2.6 Diamides of Dipicolinic Acid in Complexation and Separation of Selected Metals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .20 2.7 Recovery of Uranium by Adsorbents with Amidoxime and Carboxyl Groups: A Density Functional Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 2.8 Theoretical Studies on the Electronic Structure and Chemical Bonding of UX5–(X = F, Cl) Complexes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .23 . 2.9 First-principles Calculation of Intrinsic and Defective Properties of UO2 and ThO2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 2.10 Modeling the Autocatalytic Reaction between TcO4- and MMH in HNO3 Solution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 2.11 Fluorescent BINOL-Based Sensor for Thorium Recognition and a Density Functional Theory Investigatio . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 .
  • 2.12 Exceptional Selectivity for Actinides by N,N’-Diethyl-N,N’-Ditolyl-2,9Diamide-1,10-Phenanthroline Ligand: A Combined Hard-Soft Atoms Principle# . . . . . . . . .28 2.13 The Studies on Optimization of the Separation Method of Am and Cm . . . . . . . . . . . . .29 2.14 Burn-up Calculation of Plutonium in Fusion-fission Hybrid Reactor . . . . . . . . . . . . . . . .30 2.15 [UO2(NO3)4]2- Complex in Ionic Liquids Investigated by Optical Spectroscopic and Electrochemical Studies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 2.16 Complexation of Uranyl by Neutral Bidentate Phosphonate Ligands in Ionic Liquids . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 . Session 3: Waste Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .35 3.1 Sorption of Uranium and Rhenium in the Presence of Fulvic Acids . . . . . . . . . . . . . . . . .36 3.2 Oxalic Acid Effect on the Diffusion of Se(IV) and Re(VII) in Bentonite . . . . . . . . . . . . . . .37 3.3 Migration of Actinides and fission products in Environments . . . . . . . . . . . . . . . . . . . . . .39 3.4 Development of Negative Ce Anomalies in Biogenic Mn Oxide: the Role of Microorganism on REE Mobility during the Bio-oxidation of Mn2+ . . . . . . . . . . . . . . . . . . . . .41 3.5 New Biotechnology Methods for Radioactive Wastes Treatment . . . . . . . . . . . . . . . . . . . 42 3.6 Removal of Radioactive Cesium from Soil and Sewage Sludge Contaminated by Fukushima Daiichi NPP Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .43 3.7 Synthesis of Multifunctional Silica-based Adsorbents and Their Application in Decontamination of Radioactive Contaminated Wastewater . . . . . . . . . . . . . . . . . . . . . . . 45 . 3.8 Remove uranium and Fluorine from Wastewater . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .46 3.9 Irradiation Stability of the Tributyl Phosphate Solvent Extraction System . . . . . . . . . . . . 48 . 3.10 U(VI) Sorption on Silica in the Presence of Short Chain Mono-carboxylic Acids . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .50 3.11 Effect of Some Ions on the Sorption of Th(IV) to K-feldspar . . . . . . . . . . . . . . . . . . . . . 51 . 3.12 Uranyl Ions Sorption to TiO2 and Interaction with Sorbed FA: Experiments and Modeling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .52 3.13 Thermal Decomposition Behavior of Nitrate Solution Containing Di-nbutylephosphate in Vitrification Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .53 3.14 Study on the Synthesis of AMP Loaded Silica and Its Adsorption Behavior for Cs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .54 3.15 Selective Adsorption and Stable Solidification of Sr by Potassium Titanates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .56 3.16 Adsorption and Stable Solidification of Cesium by Insoluble Ferrocyanide Loaded Porous Silica Gels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 . 3.17 Separation of Nuclides by Different Types of Zeolites in the Presence of Boric Acid . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .58 Session 4: Transmutation, Resources and Materials Utilization, etc. . . . . . . . . . . .59 4.1 Hydriding Properties of Uranium Alloys - Their Meaning for Nuclear Fuel Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 4.2 Microstructural Study of As-Cast U-Rich U- Zr Alloys . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 .
  • 4.3 Production of Standard Particles and Their Application in Particle Analysis for Nuclear Safeguards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 4.4 Après ORIENT, A New P&T Challenge to Transmute Radioactive Wastes into Resources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 . 4.5 The Numerical Analysis about the Creation of Strategic Important Elements by Nuclear Transmutation Processes of Fission Products . . . . . . . . . . . . . . . . . . .66 Session 5: Hydro-Separation Technologies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 . 5.1 Current Status of Reprocessing Process using Pyridine Resin in Hydrochloric Acid Solution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 . 5.2 Studies on the Advanced Hybrid Reprocessing System “FluoMato” Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 5.3 R&D Efforts Using Novel Extractants for the Development of ‘Green’ Separation Technologies Relevant in the Back-End of Nuclear Fuel Cycle . . . . . . . . . . . . . .73 5.4 Preparation of High Purity Thorium by Centrifugal Extraction . . . . . . . . . . . . . . . . . . . . . 75 . 5.5 Development of Selective Separation Method for Nuclear Rare Metals Using Highly Functional Xerogel Microcapsules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 . 5.6 Novel Pillar[5]arene-Based Phosphine Oxides as Extractants for the Segregation of f-Block Elements from Acidic Media in Biphasic Systems . . . . . . . . . . . . . . .77 5.7 Synthesis and Adsorptivity of Acryloylmorpholine Resin for Selective Separation of U(VI) in Nitric Acid Media . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .79 5.8 Adsorption Behavior of Am(III) and Ln(III) from Nitric Acid Solution onto isoHexyl- BTP/SiO2 -P Adsorbent. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .80 5.9 Preparation of Anion Exchanger by Pre-irradiation Grafting Method and Its Adsorptive Removal of Rhenium as an Analogue of Radioactive Technetium . . . . . . . . . . .81 5.10 Adsorption of Th4+ from Aqueous Solution onto Poly(N,Ndiethylacrylamid e-co-acrylic acid) Microgels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .82 5.11 Recovery of 233U from Irradiated Thorium Oxide Using 5% TBP as Extractant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .85 5.12 Synthesis and Characterization of UO2 2+-ion Imprinted Polymer for Separation and Preconcentration of Trace Uranyl Ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 . 5.13 Solid Phase Extraction Using N-doped Carbonaceous Covalent Organic Frameworks for Treatment of Uranium (VI) Ions from Water Solutions . . . . . . . . . . . . . . . . .88 5.14 Extraction of Thorium(IV), Uranium(VI) and Rare Earths with NTAamide . . . . . . . . . . . 90 . 5.15 Adsorption and Separation Characteristics of Thorium from Nitric Acid Solution Using Silica-Based Anion Exchange Resins . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 . 5.16 Adsorption and Elution of Rhenium (VII) with a Porous Silica-based Anion Exchanger AR-01 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .93 5.17 Study on the Properties of isoBu-BTP/SiO2-P Adsorbent in the Separation of Minor Actinides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 5.18 Removal of Th4+ Ions from Aqueous Solutions by Graphene Oxide . . . . . . . . . . . . . . .95
  • 5.19 Influence of γ-irradiation on the isoBu-BTP/[C2mim][NTf2] Extracting System during Dy(III) Extraction . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . 96 5.20 Ethanolamine-isocyanate Modified Graphite Oxide for Selective Solid-phase Extraction of Uranium .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98 . 5.21 Separation Behavior of Rare Metals by Functional Xerogels Impregnated with MIDOA Extractant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100 Session 6: Pyro-Separation Technologies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .101 6.1 Recent Study on Pyrochemical Treatment of Spent Nitride Fuels in JAEA . . . . . . . . . . 102 . 6.2 Thorium based Molten Salt Fuel Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103 6.3 The Study on the Solubility of Rare Earth Oxides in a New Molten Salts LiCl-NaCl-MgCl2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 104 6.4 Separation of SmCl3 and DyCl3 by Galvanostatic Electrolysis in LiCl-KCl Melts at Magnesium Electrodes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .106 6.5 Electrochemical Extraction of Holmium in LiCl-KCl-HoCl3 Melts on a Nickel Electrode . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 . 6.6 Electrochemical Behaviors of REs in FLINAK Eutectics . . . . . . . . . . . . . . . . . . . . . . . . 110 . 6.7 Electrochemical Behavior of Cerium and Electrodeposition of Al–Li–Ce Alloys from Molten Chlorides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111 . 6.8 Electrochemical Extraction of Thulium in LiCl–KCl Melt Containing TmCl3 at Liquid Zn Electrodes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .112 6.9 Electrochemical Behavior of Erbium and Aluminum in the LiCl-KCl Eutectic . . . . . . . . .114 6.10 Electrochemical Extraction of Samarium from LiCl-KCl Melt by forming Sm-Zn Alloys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 116 . 6.11 Molecular Dynamics Simulation of Molten LiF-ThF4 Salt Systems . . . . . . . . . . . . . . 117 Session 7: Innovative Materials and Separation . . . . . . . . . . . . . . . . . . . . . . . . . . .118 7.1 Study on Proton Beam Irradiation of Ionic Liquid . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 7.2 Surface Modification of Carbon Nanomaterials and their Application in Radionuclide Pollution Cleanup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 121 . 7.3 Extraction Uranium from Aqueous Solution with Malonamide into Ionic Liquid . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 122 . 7.4 Extraction of Uranium(VI) and Thorium(IV) Ions from the Aqueous Phase into an Ionic Liquid by 4-oxaheptanediamides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .125 7.5 Radiation Effect on EuIII Extraction Ability of BTPhen ILs System . . . . . . . . . . . . . . . 127 7.6 Separation of Uranyl Species Using Task-specific Ionic Liquid, [Hbet][Tf2N] . . . . . . . . 129 7.7 Dissolution of UO2 in the System of [Imim][FeCl4]-DMSO . . . . . . . . . . . . . . . . . . . . . . 130 7.8 Influence of Solvent Structural Variations on the isoBu-BTP [Cnmim][ NTf2] Extracting System during Eu(III)/Dy(III) Extraction . . . . . . . . . . . . . . . . . . . . . . . . . . .132 7.9 Extraction of Several Rare-earth Metal Ions Using isoBu-BTP[C2mim][ NTf2] System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 134 .
  • 7.10 Electrodeposition of Rh(III) and Pd(II) from 1-Ethyl-3-Methylimidazolium Trifluoroacetate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 136 7.11 Adsorption of Thorium on Magnetic Multi-walled Carbon Nanotube . . . . . . . . . . . . . . 138 . 7.12 A Catechol-like Phenolic Ligand-functionalized Hydrothermal Carbon : One-pot Synthesis, Characterization and Sorption Behavior towards Uranium . . . . . . . . . .139 7.13 A Simple Approach to Highly Efficient Uranium Selective Sorbent : Preparation and Performance of a Novel Amidoxime-functionalized Hydrothermal Carbon . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .141 7.14 Amidoxime-Grafted Multiwalled Carbon Nanotubes by Plasma and its Application in the Removal of Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 143 . 7.15 Amino Functionalized MIL-101 Metal–Organic Frameworks (MOFs) for U(VI) Capture . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .144 7.16 A Novel Functionalized 2-D COF Materials : Synthesis and Application as Selective Solid-phase Extractant in Separation of Uranium . . . . . . . . . . . . . . . . . . . . . . . . .145 7.17 Comparation of Ce(IV) Stripping Rate from TBP and DBP . . . . . . . . . . . . . . . . . . . . . 147 . 7.18 Impact of Low Molecular Weight Organic Acids on Uranium Uptake and Distribution in a Variants of Mustard (Brassica juncea var.tumida) . . . . . . . . . . . . . . . . . . . 148 . 7.19 Sorption of Selenium(IV) on Modified Bentonit . .. . . . . . . . . . . . . . . . . . . . . . . . . . 149 7.20 Pyrohydrolysis of Fluorides from Thorium-based Molten Salt Reactor . . . . . . . . . . .151 7.21 Comparative Study on Sorption of Eu(III) to Two Kinds of Mica Muscovite and Phlogopite . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 152 . 7.22 Sorption of Np(V) onto Na-bentonite : Effect of equilibrium time, pH, ionic strength and temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .154 7.23 Application and Evaluation of Radioisotope in Tracer Technique. . . . . . . . . . . . . . . . .155 7.24 Extraction of U(VI) and Th(IV) from Aqueous Solution into Ionic Liquid or N-pentanol Using Methylimidazole Derivatives as Extractants . . . . . . . . . . . . . . . . . . . . . . 156 .
  • Plenary Lectures 1
  • Some Hot Issues on Nuclear Energy Chemistry in China Zhifang Chai Nuclear Energy Chemistry Group, Key Laboratory of Nuclear Analytical Techniques Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049, China E-mail: chaizf@ihep.ac.cn Nuclear energy future in China will be still bright following the Fukushima Accident. The reason is straightforward: (1) Nuclear energy, per se, is a safe and clean energy source; (2) China can not survive as a productive economy without nuclear energy, and in the meantime it needs to control the emission of the green house gas. Therefore, there is a strong impetus to develop nuclear energy in China, which is now experiencing a renaissance. In this talk, the recent achievements in nuclear energy chemistry of China are selectively highlighted, with emphasis on the extraction of uranium from seawater, front-end chemistry, actinide coordinated chemistry associated with nuclear fuel fabrication, actinide solution chemistry and nuclear fuel reprocessing. Another key issue is how to apply nano-materials and nano technology in nuclear energy chemistry. Some positive measures for promotion of the nuclear energy chemistry in China will be addressed, and future perspectives will be briefly outlined as well. Nuclear energy chemistry in China needs new thoughts, new methods and new materials; needs multidisciplinary research; and, particularly, needs bright young scientists. Acknowledgement This work was supported by Natural Science Foundation of China (Grants 91026007, 91226201 and 11275219) and the "Strategic Priority Research Program" of the Chinese Academy of Sciences (Grants XDA030104). References 1. WQ Shi, YL Zhao, ZF Chai. Radiochim Acta. 2012, 100: 529. 2
  • Present status of nuclear in Japan after the accident of Fukushima Daiichi Toshio Wakabayashi Tohoku University, Japan The great earthquake of magnitude 9 and the later tsunami on March 11, 2011 gave very serious damages to the East Japan area. About nuclear power plants of the East Japan, 11 plants were operated before the earthquake and all the plants were automatically safely stopped at the time of the earthquake. However, as for Fukushima Daiichi Nuclear Power Plant, a large quantity of radioactive materials was released by meltdown of the core and the hydrogen explosions of reactor buildings after tsunami. Many inhabitants within the area of 30km of Fukushima Daiichi nuclear power plants are now evacuating. Present status of nuclear in Japan after the accident of Fukushima Daiichi is introduced in this paper. The status of Fukushima Daiichi nuclear power station and the status of the long-and-mid term roadmap towards the decommissioning are shown as follows. Cold Shutdown Condition is maintained at Unit 1-3. Measures to complement status monitoring are being implemented. The RPV bottom temperature and the PCV gaseous phase temperatures at Units 1-3 were approx.30-50 degrees (as of October 19) and fulfill the requirement (100 degrees or less). The highly radioactive water accumulated in the building basement is treated to be used for reactor cooling. The contaminated water generated in this process treated and stored. Preparation for fuel removal from the spent fuel pool is in progress. Debris removal from the upper part of Units 3-4 Reactor Building is in progress to prepare for fuel removal from the spent fuel pool. The Nuclear Regulation Authority(NRA) was established in September 2012 to absorb and learn the lessons of the Fukushima Daiichi nuclear accident of March 11, 2011. The fundamental mission of the NRA is to protect the general public and the environment through rigorous and reliable regulations of nuclear activities. The new regulatory requirements were decided taking into account the lessons-learnt from the accident at Fukushima Daiichi Nuclear power plants. Main requirements are shown as follows. (1) Measures to prevent core damage(postulate multiple failures) (2) Measures to prevent containment vessel failure (3) Measures to suppress radioactive materials dispersion (4) Consideration of internal flooding (5) Consideration of natural phenomena in addition to earthquakes and tsunamis--volcanic eruptions, tornadoes and forest fires (6) Response to intentional aircraft crashes Concerning the reprocessing plant of the Japan Nuclear Fuel Limited(JNFL) in Rokkasho, the completion timing of the reprocessing plant is being examined based on the evaluation status of the 3
  • nuclear power station and trend of the new regulatory requirements on cycle facilities. The new process will be notified as soon as it has been organized. JNFL has been constructing the Vitrification Technology Development Facility, which is the base of research and development, within the reprocessing site in order to further improve vitrification technology. The Japan Atomic Energy Agency(JAEA) will be reformed to focus on the Fukushima Daiichi nuclear accident support, the research for enhancement of nuclear safety, the basic nuclear research, and R&D for nuclear fuel cycle including Monju development. 4
  • Session 1: General Issues on Nuclear Energy and Fuel Cycle 5
  • ENVISION OF WORLD NUCLEAR ENERGY /FUEL CYCLE DEVELOPMENT AND CHINA’s ACTION GU Zhongmao China Institute of Atomic Energy / Shanghai Jiaotong University The worldwide nuclear energy development including China after Fukushima nuclear accident is briefly viewed. The international general trend of fuel cycle for sustainable development is envisioned, and China’s efforts to develop advanced nuclear fuel cycle are described. Data shows that the global nuclear energy development has stepped out of the shadow of Fukushima accident. Advanced nuclear fuel cycle, or Fast reactor cycle, is a sustainable way of nuclear fission energy. Such understanding is becoming the consensus of the world nuclear community. China has a big nuclear energy program and must establish an advanced nuclear fuel cycle system, which is geared to the international trends. 6
  • The Role of Advanced Reprocessing Technology on 3S (Safety, Security, and Safeguards) in Nuclear Fuel Cycle and Radioactive Waste management Jor-Shan CHOI 1 1 UC Berkeley Nuclear Research Center, University of California at Berkeley, CA, USA, Email: jorshan@yahoo.com, jorshan@nuc.berkeley.edu ABSTRACT: In the IAEA “Milestones in the Development of a National Infrastructure for Nuclear Power”, the importance of nuclear safety, security, and safeguards /nonproliferation (3S) in the peaceful use of nuclear energy was recognized. In 3S, nuclear safety deals with the prevention and mitigation of nuclear accidents and the release of radioactivity; nuclear security deals with the prevention and detection of and response to the theft, sabotage, unauthorized access, illegal transfer, or other malicious acts involving nuclear and radiological materials or their associated facilities; and nuclear safeguards/nonproliferation deals with the prevention of the spread of nuclear weapons, or materials used in fabricating such weapons. In the aftermath of the Fukushima accident in March 2011, issues associated with managing and disposing of used nuclear fuel moved “front-and-center”. The event exposed the safety concern in prolong storage of used fuel in water pool, it also highlighted the intractably technical, institutional, and societal problems in used-fuel management. Used fuel contain the radioactivity which if released, could cause a widespread radiological consequences and environmental contamination. They also contain materials (i.e., unfissioned235U and plutonium) that if separated, are the aspired targets for terrorists, and perhaps even for the host countries producing such materials for use in improvised or stockpiled nuclear devices. Thus, used-fuel management involving advanced reprocessing technologies has all the characteristics of 3S. Advanced reprocessing technology employing pyro-processing recovers from used fuel the transuranic that contains plutonium; minor actinide (i.e., neptunium, americium, curium); and a small percentage of lanthanide for recycling in future metal-fuel fast reactors. The pyro technology is advocated as proliferation resistant because plutonium is not cleanly separated. The advanced aqueous process based on selective adsorption technology aims to separate plutonium and minor actinide cleanly for recycling in existing LWRs and future fast reactors. The aqueous technology is advocated as beneficial to radioactive waste management. The questions of “what is the motivation for used-fuel treatment technologies?” and “what is the role which advanced reprocessing technologies can play in nuclear fuel cycle and radioactive waste management?” would be assessed here, in the context of the nuclear 3S. KEYWORDS: 3S, nuclear safety, security, safeguards, used fuel, advanced reprocessing technologies, pyro-processing, aqueous process based on selective adsorption technology. Viewpoints expressed here are those of the author and not necessarily those of his affiliation. 7
  • Flexible Fuel Cycle Initiative to Cope with the Uncertainties after Fukushima Daiichi NPP Accident Tetsuo Fukasawa Hitachi-GE Nuclear Energy, Ltd. 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 Japan, Tel: +81-294-55-4319, Fax: +81-294-55-9904 E-mail: tetsuo.fukasawa.gx@hitachi.com Fast breeder reactors (FBR) nuclear fuel cycle is needed for long-term nuclear sustainability while preventing global warming and maximum utilizing the limited uranium (U) resources. The “Framework for Nuclear Energy Policy” by the Japanese government on October 2005 stated that commercial FBR deployment will start around 2050 under its suitable conditions by the successive replacement of light water reactors (LWR) to FBR [1]. Even after Fukushima Daiichi Nuclear Power Plant accident which made Japanese tendency slow down the nuclear power generation activities, Japan should have various options for energy resources including nuclear, and also consider the delay of FBR deployment and increase of LWR spent fuel (LWR-SF) storage amounts. As plutonium (Pu) for FBR deployment will be supplied from LWR-SF reprocessing and Japan will not possess surplus Pu, the authors have developed the flexible fuel cycle initiative (FFCI) for the transition from LWR to FBR [2]. This FFCI system is also effective after the Fukushima accident for the reduction of LWRSF and future LWR-to-FBR transition. The outline of FFCI shown in Fig. 1 consists of U removal as LWR-SF reprocessing and Pu+U(+MA) recovery as reprocessing of U removal residue (recycle material, RM) and FBR-SF. The U removal residue has less than 1/10 of the LWR-SF amounts and higher Pu concentration with FP, which enables the compact interim Pu storage with high proliferation resistance and compact Pu+U(+MA) recovery just before FBR use. Removed U is easily re-enriched after purification for LWR reuse. MA would be recovered from stored RM after the development of partitioning and transmutation technology. In this work, the amounts of Pu, reprocessing, LWR-SF were calculated and compared for the FFCI and the ordinary cycle with full LWR/FBR-SF reprocessing, which revealed that the FFCI could supply enough Pu and no excess Pu to FBR in any cases. [1] Atomic Energy Commission of Japan, “Framework for Nuclear Energy Policy”, October 11, 2005. [2] T. Fukasawa, et al., “Flexible LWR-to-FBR Transition Fuel Cycle System”, Proc. GLOBAL 2011, No. 355737, Makuhari, Japan, December 11-16, 2011. LWR Spent fuel Most U removal RM Storage Fresh Pu+U(+MA) fuel recovery FBR Recovered U Fresh fuel U Storage FP(+MA) Spent fuel RM: Recycle Material, most U removal residue which contains Pu+U+MA+FP MA: Minor Actinides, Np+Am+Cm; FP: Fission Products Fig. 1 The outline of Flexible Fuel Cycle Initiative (FFCI) system 8
  • Advanced Fuel Cycle : Status and Technology Development at KAERI J.H. Leea,*, H.S. Leeb,c,*, J.W. Leec, J.M. Hurc, J.K. Kimc, S.W. Paekc, I.J. Choc, W.I. Koc, I.T. Kimc, G.I. Parkc and H.D. Kimb, c a Department of Nanomaterials Engineering, and bGraduate School of Green Energy Technology, Chungnam National University, 79 Daehak-ro, Yuseong-gu, Daejeon 305-764, Republic of Korea c Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 1045 Daedukdaero, Yuseong, Daejeon 305-353, Republic of Korea * Corresponding author: jonglee@cnu.ac.kr, hslee5@kaeri.re.kr Pyroprocessing technology has been actively developed at Korea Atomic Energy Research Institute (KAERI) to meet the necessity of addressing spent fuel management issue. This technology has advantages over aqueous process such as less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, and compact equipments. This presentation describes the pyroprocessing technology development at KAERI from head-end process to waste treatment as well as safeguards R&D. The unit process with various scales has been tested to produce the design data associated with scale-up and selected data will be presented in this presentation. Pyroprocess integrated inactive demonstration facility (PRIDE) was constructed at KAERI and it began test operation in 2012. The purpose of PRIDE is to test the process regarding unit process performance, remote operation of equipments, integration of unit processes, scale-up of process, process monitoring, argon environment system operation, and safeguards-related activities. The test of PRIDE will be promising for further pyroprocessing technology development. Fig. 1. Exterior of PRIDE (left) and Bird’s-eye view of argon cell (right) 9
  • The Nuclear Education and Training Program at University of California Irvine Mikael Nilsson1*, George Miller2, A.J. Shaka2 1 Department of Chemical Engineering and Materials Science, 2 Department of Chemistry, University of California Irvine, Irvine, CA 92697-2575 * Corresponding author: nilssonm@uci.edu As we project into the future it is clear that the demand for energy, and especially clean energy, will rise. Concerns about rising CO 2 levels in the atmosphere have turned many eyes back again towards nuclear energy. In the last few years the interest in nuclear energy has increased not only in the US but in other parts of the world. In spite of unfortunate incidents, issues with nuclear power plants and their siting appear to be solvable with future generation reactor designs, and better attention to siting requirements. One issue that clearly needs research and development is the handling of nuclear materials both in preparation of new fuels and in handling spent fuels. The result is that the demand for personnel with the right type of training is increasing. Furthermore, recent events that have received much attention in the media surrounding the nuclear power plants in southern California is a clear indication that education, and particularly education of the public, in this region is needed now more than ever. At the University of California Irvine our Nuclear Group has, in the last few years, focused on training and research in the critical associated fields of radiochemistry, nuclear chemical engineering and nuclear materials. A previous radiochemistry program existed [1] and although most of the faculty from that time are gone the infrastructure remains, including a 250kW TRIGA reactor, which serves as the flagship of our program. Our current program includes 6 full-time faculty and staff members, 12-15 graduate students and 8-10 undergraduate students all involved in nuclear science research. The number of students involved has grown from none in 2008 to around 25 graduate and undergraduate students in 2013. The student demographics in our program consist of chemical engineering, materials science engineering, and chemistry majors making the current emphasis of our program on radiochemistry and nuclear chemistry. To strengthen our mission, UCI recently became part of the SUCCESS PIPELINE nuclear science security consortium [2], a group funded by the National Nuclear Security Administration (NNSA) to work on issues broadly related to nuclear security. This consortium has as its primary goal to ensure that there is a nuclear science educated workforce in the US. Collaboration with minority serving institutes is highly encouraged so that individuals from all backgrounds can have an opportunity to be included in the future nuclear science workforce. Within our program there are ample opportunities for collaborations and internships in the areas of nuclear energy (including reactor operations, and instrumentation), nuclear medicine, environmental remediation studies, and nuclear forensics. Please contact the authors with inquires about our program. [1]. V.P. Guinn, G.E. Miller, F.S. Rowland. Radiochemistry teaching and research at UC Irvine, Nucl. Technol., 27, 1, 124, (1975). [2]. http://nssc.berkeley.edu/ (Accessed Oct 14, 2013) 10
  • Session 2: Basic Chemistry of Actinides and Fission Products 11
  • Utilization of Technetium and Actinide Compound Synthesis and Coordination Chemistry for the Nuclear Fuel Cycle: Exploring Separations, Fuels, and Waste Forms K.R. Czerwinskia, A. Bhattacharyyaa, J. Droesslera, W. Kerlina, E. Johnstonea, F. Poineaua, P. Wecka, E. Kima, P. Forstera, T. Hartmanna, and A. Sattelbergera,b, a Radiochemistry Program, University of Nevada, Las Vegas, Las Vegas, Nevada, USA b Energy Engineering and Systems Analysis Directorate, Argonne National Laboratory * Corresponding author: czerwin2@unlv.nevada.edu Radiochemistry is a discipline that explores chemical and nuclear properties of elements and their isotopes. Within radiochemistry technetium and the actinides elements are unique in that they lack stable isotopes. These radioelements are germane to nuclear technology and also represent an underexplored section of the periodic table. The actinide elements compose the fuel in reactors and are produced from neutron capture. In the nuclear fuel cycle technetium has a unique role. It is produced at a significant level and is an important fission product for waste consideration. Compared to other elements on the periodic table, technetium and the actinides is less explored, especially in areas of compound synthesis and coordination chemistry. The nuclear fuel cycle offers opportunities to investigate fundamental and applied technetium and actinide chemistry in more detail, with fundamental complexation chemistry providing insight into waste forms, fuels, and separations. Examples are given for technetium and actinide solution and solid phases, with the coordination chemistry explored by spectroscopy and diffraction. An overview on technetium waste forms is provided, highlighting the need for fundamental information on this element to improved synthetic routes and understand resulting behavior. The thermal and hydrothemal based synthesis of technetium compounds is described. Spectroscopic and diffraction results are provided. Trends in the products from computation [1] and experiment are discussed, emphasizing the role of technetium-technetium interaction with oxidation state change. For waste forms, low valent or metallic phase formation demonstrates enhanced inter-technetium interactions which grants the resulting compounds resistance to corrosion or limits solubility. Development of advanced fuels can leverage innovative synthetic techniques that are utilized in the laboratory and non-nuclear industry. In particular methods that use novel reactions with common starting materials can be applied to produce fuels with suitable attributes for advanced fuel cycles. An example is provided based on the formation of uranium mononitride from dinitride starting material [2]. Uranium dinitride is air stable and can be produced from oxide starting material. Uranium dinitride pellets can be formed in air and then sintered under inert atmosphere to produce uranium mononitride. The unique method for the nitride synthesis can be coupled with established sintering techniques to produce fuel. These waste form and fuel illustrations exemplify the utility synthesis reactions can play in the future fuel cycles. A final example is provided on the utility of radioelement synthesis and coordination chemistry in solutions. In one case the use of ionic liquids as a novel media for nuclear separations is presented, emphasizing electrochemistry of the actinides. Understanding the dissolution chemistry and potentials of electrodeposition for actinides and lanthanides in the tri-methyl-n-butyl ammonium n-bis(trifluoromethansulfonylimide) ([Me3NBu][TFSI]) ionic liquid is explored. Studies of the species in the ionic liquid using UV-Visible and X-ray spectroscopy have been performed, along with electrochemistry studies and scanning electron microscopy examination of deposited phases. A method of direct dissolution is currently being investigated and has been successful for a uranium oxide and lanthanide carbonates [ 3,4]. Determining the mechanism of uranium dissolution is the near term 12
  • goal of this research. Initial data support the conclusion that the dissolved uranium species in the ionic liquid is UO22+ .The equatorial coordinating oxygens could be from the small amount of water present or the sulfonyl on the [TFSI] anion. The cyclic voltammetry of U in [Me3NBu][TFSI] shows that the system can support investigation of 5 V potential windows. The electrochemistry also shows a complex series of peaks for U in [Me3NBu][TFSI]. From the initial results the examined ionic liquid system provides the necessary components to provide separations of actinides and lanthanides from spent nuclear fuel. Solution based separation of trivalent lanthanides from Am and Cm is also provided as an example of the utility of speciation and coordination chemistry in the nuclear fuel cycle. Soft donor ligands such as dithiophosphinic acids and bis-1,2,4-triazinylpyridine/bipyridine (BTP/BTBP) derivatives show significant separation selectivity. Many of these ligands are limited by poor stability, constrained working pH range, solubility in suitable solvents, and competition from counter anions. Various triazinyl and bis-triazinylpridine (H, Methyl, Ethyl, Pyridyl and Phenyl) derivatives have been synthesized and their complexation with Eu3+, Tb3+ and Cm3+ by time resolved laser fluorescence spectroscopy presented. The solvent is found to play a significant role in the complexation behavior and resulting speciation and coordination. In the acetonitrile medium, the complexes contain one ligand molecule per metal ion. Spectroscopic signatures change to ML3 species in methanol medium. For hard acceptors acetonitrile is known to be less solvating as compared to methanol. The Eu3+ ion, being a hard cation, is less solvated by acetonitrile and the nitrate counter anion strongly binds with it and the BTP molecules. When the Eu(III) complex of Py-BTP was prepared in acetonitrile medium, the single crystal XRD result shows that it acts as a tetra-dentate ligand with the stoichiometry Eu(Py-BTP)(NO3)3 resulting in 10 coordinated Eu(III) ion. The overall results show the utility of radioelement speciation, compound synthesis, and coordination chemistry in expanding general chemistry knowledge and the development of applications exploiting radionuclide synthesis, speciation, and coordination chemistry. 1. Weck, P.F., Kim, E., Poineau, F., Rodriguez, E.E., Sattelberger, A.P., Czerwinski, K.R. Inorg. Chem. 48(14), 6555-6558 (2009). 2. Yeamans, C.B., Silva, G.W.C, Cerefice, G.S., Czerwinski, K.R., Hartmann, T., Burrell, A.K., and Sattelberger, A.P. J. Nucl. Mat. 347, 75-78 (2008). 3. Hatchett, D.W., Droessler, J., Kinyanjui, J.M., Martinez, B., Czerwinski, K.R. Electrochim. Acta, 89, 144-151 (2013). 4. Pemberton, W.J., Droessler, J.E., Kinyanjui, J.M., Czerwinski, K.R., Hatchett, D.W. Electrochim Acta., 93, 264-271 (2013). Figure 2. liquid. Figure 1. EXAFS data showing uranyl in the ionic The data show the formation of oxidized uranium from species dissolution. Computation study on Tc halides showing difference with the iodine system. 13
  • Heptavalent State of Transuranium Elements, Technetium and the Other Elements of the Periodic Table K.E. Germana,*, K. Czerwinskib, M.S. Grigorieva, A.V. Safonov a, F. Poineaub, V.F. Peretrukhina a A.N. Frumkin Institute of Physical Chemistry and Electrochemistry RAS 31, Leninsky prospekt, Moscow, 119071, Russia. * - guerman_k@mail.ru b University of Nevada Las Vegas, LasVegas, USA The discovery of new compounds where transuranic elements are present in heptavalent oxidation state (1967, 1974 [1, 2]) has been the front point for its identification as a more complicated (relative to lanthanides) group in the Periodic table. This observation initiated profound comparison of these compounds to the elements of the 4th – 6th periods. Simulteneously, it formed a critical view at the limitations that were prescribed to the lighter elements in their highest oxidation states. In the transition from one chemical element to the next at the beginning of each period of the Periodic Table the maximum oxidation state of the elements monotonically increases from one to seven, after that only Ru and Os in the 5th and 6th periods continue this pattern, being oxidized up to octavalent state. In all other periods heptavalent elements are followed by elements of lower than +7 maximum valences. The heptavalent state of Np, Pu and Am was unforeseen by the actinide conception. Its discovery led to series of discussions about the similarities and differences between properties of the heptavalent transuranic elements (TRU) and elements of Group VII of the short form of the Periodic Table, indicating the need for further development of the actinide conception [3,4]. Current work continues the discussion with the use of data on the crystal structure of the new compounds of heptavalent elements published in recent years. Halogenide(VII) derivatives and heptavalent d-elements (Mn, Tc, Re) have many similarities in structure and properties. However they have some interesting differences concerning not only the redox properties, which is quite evident and understandable, but also unexpected differences in the composition and properties of their crystalline hydrates [5,6], the structure of the oxides ([7]) and acids such as [TcO 3 (OH)(H 2 O) 2 ] [8-10]. TcO 3 (H 2 O) 2 (OH) Na4 [AnO 4 (OH) 2 ](OH)∙2H 2 O [13] [10] Solid compounds of transuranic elements (VII) are obtained up to date for Np and Pu, both by solid phase reactions, and from aqueous alkaline solutions. Each year, several new compounds of heptavalent TUE are synthesized and their crystal structure is determined. TRU(VII) compounds are varied in composition and can be regarded as containing anions AnO 6 5-, AnO 5 3-, [AnO 4 (OH) 2 ]3-, [An 2 O 8 (OH) 2 ]4- and AnO 4 - [11-17]. For a number of compounds formally containing the first two types of anions the isostructurality to the corresponding ortho- and mesorhenate was established. Unlike them, compounds formally containing anions AnO 4 -, are not analogues of compounds with such a composition, formed by the elements of the seventh group of the Periodic table, and in fact they do not contain single-charged tetraoxide anions. As it was established by one of us earlier, compounds of the type MAnO 4 (·nH 2 O) (M - alkali metal) are analogs of alkaline-earth metal uranate (VI). They contain shortened linear groups AnO 2 3-, combined by bridging O atoms in the anionic layers [13]. Presence of the linear [O = An = O] groups in the crystal structure of Np, Pu, Am compounds again indicates that the "yl" group is a characteristic feature of transuranic compounds in 14
  • higher oxidation states V, VI, VII and at the same time does not appear in the structure of the compounds of heptavalent halogens and heptavalent Mn, Tc, Re (being present just in several Tc(V)O 2 + complexes). Compounds with anions [AnO 4 (OH) 2 ]3- are synthesized from solutions in the form of monocrystals. Recently, systematic studies on the synthesis and X-ray analysis of Np(VII) and Pu(VII) compounds of such a type have been carried out. Crystal structures of a range of Np(VII) compounds previously defined were specified. About 20 new compounds of An(VII) were synthesized, for the first time including 10 Pu(VII) compounds in the form of single crystals, their crystalline structures were defined. Generally Pu(VII) compounds are isostructural to the corresponding Np(VII) compounds, which confirms the chemical similarity of heptavalent neptunium and plutonium. Among the synthesized and examined compounds there are compounds of two new types: mixed-cationic containing two different alkali metals and Na 4 [AnO 4 (OH) 2 ](OH)·2H 2 O (An = Np, Pu) compounds containing outer OH-groups. [AnO 4 (OH) 2 ]3anions form tetragonal bipyramide in which OH-groups are in apical positions at distances An-O ~ 2.3-2.4 Å, and An= O distance in almost perfectly symmetrical square AnO 4 is ~ 1.9 Å. Distances An = O in AnO 4 groups change slightly from Np(VII) to Pu(VII). At the same time, there is a significant shortening of AnO (OH) bonds. Thus, actinide contraction in the Np(VII) and Pu(VII) compounds is anisotropic. For the first time it was found out that [AnO 4 (OH) 2 ]3- anions can occupy general positions, the orientation of OH-groups differing significantly from centrosymmetric one. Data obtained in recent years on the crystal structure of the new compounds of heptavalent neptunium and plutonium, pertechnetate and perrhenate confirm the earlier prevailing opinion [11] about the absence of a deep similarity in physico-chemical properties between the heptavalent transuranic elements and the elements of Group VII of the short form of the Periodic table and the formal nature of some of the structural similarities among the considered heptavalent compounds. References. 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14. Крот Н.Н., Гельман А.Д. Докл. АН СССР, 1967, т.177. № 1. С. 124-126. Крот Н.Н., Шилов В.П., Николаевский В.Б., Пикаев А.К., Гельман А.Д. Докл. АН СССР, 1974, т.217. № 3. С. 589-592. Keller C., Seiffert H. Inorg.Nucl. Chem. Letters, 1969, vjl.5, p.1205-1208. Крот Н.Н., Гельман А.Д., Спицын В.И. ЖНХ, 1969, т. 14, с. 2633-2637. A.Ya. Maruk, M.S. Grigor’ev, K.E. German. // Russian Journal of Coordination Chemistry, 2011, Vol. 37, No. 6, pp. 444–446. Герман К.Э., Крючков С.В., Беляева Л.И. // Известия АН ССР - Сер.хим. 1987, № 10, стр. 2387. Rard, J.A., Rand, M.H., Andregg, G., Etc. Chemical Thermodynamics, Vol. 3. Chemical Thermodynamics of Technetium (M.C.A. Sandino and E. Östhols, eds.), OECD Nuclear Energy Agency, Data Bank, Issy-les-Moulineaux, France 1999, 567 p. F. Poineau, Ph. Weck, K. German, K. Czerwinski etc. // Dalton Trans. (2010) 39 (37), pp. 8616-8619. K. German, A. Maruk, F. Poineau, Ph. Weck, G. Kirakosyan, V. Tarasov, K. Czerwinski. In: 7th International Symposium on Technetium and Rhenium – Science and Utilization – Book of Proceedings - July 4 -8, 2011, Moscow, Russia (Eds.: K.E. German, B.F. Myasoedov, G.E. Kodina, A.Ya.Maruk, I. D. Troshkina). Publishing House GRANITSA, Moscow 2011, p. 99-100. F. Poineau, B. P. Burton-Pye., A. Maruk, K.Czerwinski et al. Inorganica Chimica Acta, 2013, vol. 398, p. 147–150. Krot, N. N., Gel’man, A. D., Mefod’eva, M. P., Shilov, V. P., Peretrukhin, V. F., Spitsyn, V. I.: Semivalentnoe Sostoyanie Neptuniya, Plutoniya, Ameritsiya. Nauka, Moscow (1977) [in Russian]. English translation: The heptavalent state of neptunium, plutonium and americium, UCRL TRANS11798 (VAAP/SA-81/27), LLNL, Livermore (CA 94550) (1981). Grigoriev M. S., Krot N. N. Plutonium Futures “The Science” 2008. Dijon, France, 7-11 July, 2008. Abstracts Booklet. P. 282-283. Grigoriev M. S., Krot N.N. Acta Crystallogr. Sect. С: Crystal Structure Communications. 2009. V. 65, N 12. P. i91-i93. Krot N. N., Charushnikova I. A., Grigoriev M. S. Actinide contraction in compounds of oxygenated actinide ions. XVIII Менделеевский съезд по общей и прикладной химии. Москва, 23-28 сентября 2007 г. Тезисы докладов. Т. 5. С. 299. 15
  • DFT Study on a Trivalent Uranium Complex Promoted Functionalization of Carbon Dioxide and Carbon Disulfide Dongqi Wang1,*, Zhifang Chai1, Wanjian Ding2, Weihai Fang2 1 CAS Key Laboratory of Nuclear Radiation and Nuclear Energy Techniques, Institute of High Energy Physics, Chinese Academy of Sciences, Beijing, 100049 2 College of Chemistry, Beijing Normal University, Beijing, 100875 dwang@ihep.ac.cn We report a DFT mechanistic study on the functionalization of CO2 and CS2 promoted by a trivalent uranium complex (Tp*)2UCH2Ph. In the calculations, the uranium atom is described by a quasi-relativistic 5f-in-core ECP basis set (LPP) developed for the trivalent uranium cation, which was qualified by the calculations with a quasi-relativistic small core ECP basis set (SPP) for uranium. According to our calculations, the functionalization proceeds in a stepwise manner, and the CO2 or CS2 does not interact with the central uranium atom to form a stable complex prior to the reaction due to the steric hindrance from the bulky ligands but directly cleaves the U−C (benzyl) bond by forming a C−C covalent bond. The released coordination site of uranium is concomitantly occupied by one chalcogen atom of the incoming molecule and gives an intermediate with the uranium atom interacting with the functionalized CO2 or CS2 in an η1 fasion. This step is followed by a reorientation of the (dithio)carboxylate side chain of the newly formed PhCH2CE2−(E = O, S) ligand to give the corresponding product. Energetically, the first step is characterized as the rate-determining step with a barrier of 9.5 (CO2) or 25.0 (CS2) kcal/mol, and during the reaction, the chalcogen atoms are reduced, while the methylene of the benzyl group is oxidized. Comparison of the results from SPP and LPP calculations indicates that our calculations qualify the use of an LPP treatment of the uranium atom toward a reasonable description of the model systems in the present study. Reference: [1] E. M. Matson, W. P. Forrest, P. E. Fanwick, S. C. Bart, J. Am. Chem. Soc. 2011, 133: 4948. [2] W. Ding, W. Fang, Z. Chai, D. Wang, J. Chem. Theory Comput. 2012, 8, 3605-3617. 16
  • Using Phosphonates to Probe Structural Differences Between the Transuranium Elements and Their Proposed Surrogates Juan Diwua,*, Thomas E. Albrecht-Schmittb a School of Radiation Medicine and Protection (SRMP) and School of Radiological and Interdisciplinary Sciences (RAD-X), Soochow University, Suzhou, Jiangsu 215123, China b Department of Chemistry and Biochemistry, Florida State University, 102 Varsity Way, Tallahassee, Florida 32306-4390, USA * Corresponding author: diwujuan@suda.edu.cn Transuranium elements, especially plutonium, play a special role in advanced technological societies. However, owing to their radioactivity and toxicity, the related research is severely restricted. One of the outcomes of this is the use of less toxic and less or non-radioactive surrogates for transuranium elements. These include early transition metals, especially Zr4+, lanthanides (e.g. Ce4+ and Eu3+), and the early actinides, thorium and uranium. The most central question is: do these surrogates actually mimic the chemistry of transuranics? In this work, we focused on the actinide diphosphonate system, for their importance in nuclear remediation and actinide separation processes, to answer the aforementioned question. Recently, we have crystallized trivalent, tetravalent and hexavalent transuranic diphosphonate compounds as well as their surrogates. In the trivalent series, plutonium and americium compounds were synthesized. In the tetravalent series, Ce4+ and Pu4+ were mainly explored, along with Th4+, U4+ and Np4+. The structural types vary from zero-dimensional clusters, one-demensional chains, to three-dimensional frameworks. PuO22+ phenylenediphosphonate is the only transuranic hexavalent compound that we were able to synthesize. There are a number of uranyl phases that can be compared to. Additionally, in order to study the interaction between different elements, experiments of mixing Np4+ and Pu4+ with both each other and with Ce4+ or UO22+ were conducted, which yielded both ordered and disordered heterobimetallic 4f/5f and 5f/5f phosphonates. In most of the series, significant differences are found between transuranium elements and their surrogates. There are examples where isostructural series exist, but transuranium elements still have their unique properties, which are not mimicked by the surrogates. Reference: J. Diwu and T. E. Albrecht-Schmitt, in Metal Phosphonates, chapter 19, transuranium phosphonates (Eds: Abraham Clearfield, and Konstantinos Demadis), RSC Publishing, London, 2011. 17
  • From Thorium to Curium: Unprecedented Structures and Properties in Actinide Borates Shuao Wanga,*, Evgeny V. Alekseevb, Thomas E. Albrecht-Schmittc a School of Radiation Medicine and Protection (SRMP) and School of Radiological and Interdisciplinary Sciences (RAD-X), Soochow University, Suzhou, Jiangsu 215123, China b Institute for Energy and Climate Research (IEK-6), Forschungszentrum Jülich GmbH, 52428 Jülich, Germany c Department of Chemistry and Biochemistry, Florida State University, 102 Varsity Way, Tallahassee, Florida 32306-4390, USA * Corresponding author: shuaowang@suda.edu.cn The use of molten boric acid as a reactive flux for synthesizing actinide borates has been developed in the past two years providing access to a remarkable array of exotic materials with both unusual structures and unprecedented properties. [ThB5O6(OH)6][BO(OH)2]·2.5H2O possesses a cationic supertetrahedral structure and displays remarkable anion exchange properties with high selectivity for TcO4−.[1-3] Uranyl borates form noncentrosymmetric structures with extraordinarily rich topological relationships.[4-5] Neptunium borates are often mixed-valent and yield rare examples of compounds with one metal in three different oxidation states (Fig. 1). [6-7] Plutonium borates display new coordination chemistry for trivalent actinides.[8] Finally, americium and curium borates show a dramatic departure from plutonium borates (Fig. 2), and there are scant examples of families of actinides compounds that extend past plutonium to examine the bonding of later actinides.[9-12] There are several grand challenges that this work addresses. The foremost of these challenges is the development of structure-property relationships in transuranium materials. A deep understanding of the materials chemistry of actinides will likely lead to the development of advanced waste forms for radionuclides present in nuclear waste that prevent their transport in the environment. This work may have also uncovered the solubility-limiting phases of actinides in some repositories such as the Waste Isolation Pilot Plant (WIPP), and allows for measurements on the stability of these materials. [1] S. Wang, E. V. Alekseev, J. Diwu, W. Casey, B. Phillips, W. Depmeier, T. E. Albrecht-Schmitt, Angew. Chem. Int. Ed., 2010, 49, 1057-1060 [2] S. Wang, P. Yu, B. A. Purse, M. J. Orta, J. Diwu, W. H. Casey, B. L. Phillips, E. V. Alekseev, W. Depmeier, D. T. Hobbs, T. E. Albrecht-Schmitt, Adv. Funct. Mater., 2012, 22, 2241–2250 [3] P. Yu, S. Wang, E. V. Alekseev, W. Depmeier, T. E. Albrecht-Schmitt, B. Phillips, W. Casey, Angew. Chem. Int. Ed., 2010, 49, 5975-5977 [4] S. Wang, E. V. Alekseev, W. Depmeier, T. E. Albrecht-Schmitt, Chem. Commun., 2011, 47, 10874-10885 [5] S. Wang, E. V. Alekseev, J. Ling, G. Liu, W. Depmeier, T. E. Albrecht-Schmitt, Chem. Mater., 2010, 22, 2155-2163 [6] S. Wang, E. V. Alekseev, J. Ling, S. Skanthakumar, L. Soderholm, W. Depmeier, T. E. Albrecht-Schmitt, Angew. Chem. Int. Ed., 2010, 49, 1263-1266 18
  • [7] S. Wang, E. V. Alekseev, W. Depmeier, T. E. Albrecht-Schmitt, Chem. Commun., 2010, 46, 3955-3957 [8] S. Wang, E. V. Alekseev, W. Depmeier, T. E. Albrecht-Schmitt, Inorg. Chem., 2011, 50, 2079-2081 [9] M. J. Polinski, D. J. Grant, S. Wang, E. V. Alekseev, J. N. Cross, E. M. Villa, W. Depmeier, L. Gagliardi, T. E. Albrecht-Schmitt, J. Am. Chem. Soc., 2012, 134, 10682-10692 [10] M. J. Polinski, S. Wang, E. V. Alekseev, W. Depmeier, G. Liu, R. G. Haire, T. E. Albrecht-Schmitt, Angew. Chem. Int. Ed., 2012, 51, 1869-1872 [11] M. J. Polinski, S. Wang, E. V. Alekseev, W. Depmeier, T. E. Albrecht-Schmitt, Angew. Chem. Int. Ed., 2011, 50, 8891-8894 Fig. 1. A view of crystal structure of neptunium borates with three oxidation states of neptunium Fig. 2.The synthesis schemes of trivalent actinide borate compounds and the photo showing the product crystals 19
  • Diamides of Dipicolinic Acid in Complexation and Separation of Selected Metals Alena Paulenovaa*, Joseph Lapkaa, Vasiliy Babainb, Mikhaliy Alyapyshevb, Jack D. Lawc a Oregon State University, Corvallis, OR, USA Khlopin Radium Institute, St Petersburg, Russia c Idaho National Laboratory, Idaho Falls, ID, USA b * Corresponding author: alena.paulenova@oregonstate.edu. Diamides have undergone significant studies as possible ligands for use in the partitioning of trivalent minor actinides and lanthanides.[1-2] Recent research has led to the development of new nitrogen-containing reagents and methods with significant potential for accomplishing separation of trivalent metals from waste process solutions such as substituted malonic acid diamides derivatives (DIAMEX) and tetra-alkyl-diglycolamides (TODGA).[3] Substituted diamides of dipicolinic acid are of interest due to their pyridine nitrogen in proximity to the carbonyl allowing it to possibly participate in coordination. Previously it was reported that among other dipicolinamides, N,N’-N,N’-ditolyldipicolinamide (EtTDPA) shows the best extractability toward americium with a slight extraction preference over europium.[4] It is known that the addition of the bulky hydrophobic anion like chlorinated cobalt dicarbollide (CCD) tend to increase the extraction of metals by neutral ligands. Many ligands were studied as synergistic additives to CCD. For CCD-based systems lanthanides and Am distribution ratios are usually close to each other, but for some poly-nitrogen compounds in the presence of dicarbollide very high separation factors can be achieved.[5-6] In our previous works the extraction ability of diamides of dipicolinic acid (DPA) in the presence of CCD was studied. It was found that DPA-CCD system selectively extract Am over lanthanides from 1-5 M nitric acid with high separation factors of Am from light lanthanides values (La-Gd). The selectivity of extraction tend to decrease with increasing of metal atomic number: DAm/DLa is > 100; while DAm/DEu does not exceed 4.[7] Understanding the underlying thermodynamic parameters of the metal:ligand interaction can lead to better ligand design for separation purposes. Small changes in the structure can affect the ability of a ligand to coordinate with metal ions in solutions. One method of determining homogenous phase binding constants is to measure the changes in absorbance during titration by UV-Vis spectroscopy. The diamides used in this study (EtTDPA) differ in only the position of the methyl group on the exterior aromatic rings yet display different affinities for varying metal oxidation states. These isomers also exhibit varying behavior within a given cation valency as well. The current work attempts to quantify the thermodynamic parameters of complexation of the trivalent lanthanide neodymium with diamides of dipicolinic acid. Figure 1: Structure of EtTDPA isomers 20
  • Diamides such as EtTDPA are neutral ligand extractants which require a balance of charge to the extracted metal cation. In the case of nitric acid the counter charge is provided by the nitrate ion, giving the mechanism of extraction: M3+ + 3NO3ˉ + nEtTDPA  nEtTDPA.M(NO3)3 where overbars indicate species contained in the organic phase. CCD exists in the organic phase of polar diluents as the acidic HCCD form. The extraction of cations is indirectly provided by CCD, acting as a charge balancer in the organic phase during a liquid-liquid cationic exchange mechanism [7]: M3+ + xHCCD  M(CCD)x(3-x)+ + xH+ The overall sum of these two equations can then be written as: M3+ + (3-x)NO3ˉ + xHCCD + nEtTDPA  nEtTDPA.M(NO3)3-x(CCD)x + xH+ The N,N’-diethyl-N,N’-ditolyl-dipicolinamides (EtTDPA, Fig. 1) were synthesized by the reaction of thionyl chloride with 2,6-pyridinedicarboxylic acid (dipicolinic acid). The acyl chloride was then reactedwith the desired isomer of N-ethyltoluidine to produce the desired EtTDPA molecule.[7] The purities of the synthesized ligands were checked by elemental analysis. The stability constants of the metal-ligand complexes formed between different isomers of N,N’-diethyl-N,N’-ditolyl-dipicolinamide (EtTDPA) and trivalent neodymium in acetonitrile were determined by spectrophotometric and calorimetric methods. Each isomer of EtTDPA was found to be capable of forming three complexes with trivalent neodymium, Nd(EtTDPA), Nd(EtTDPA) 2, and Nd(EtTDPA)3. Values from spectrophotometric and calorimetric titrations were within reasonable agreement with each other. The order of stability constants decrease in the order Et(m)TDPA > Et(p)TDPA > Et(o)TDPA. The obtained values are comparable to other diamidic ligands obtained under similar system conditions and mirror previously obtained solvent extraction data for EtTDPA at low ionic strengths. [1] Serrano-Purroy, D.; Baron, P.; Christiansen, B.; Glatz, J. P.; Madic, C.; Malmbeck, R.; Modolo, G.. Sep. Purif. Technol., 45, (3) 157-162 (2005) [2] Zhu, Z. X.; Sasaki, Y.; Suzuki, H.; Suzuki, S.; KIMURA, T. Anal. Chim. Acta., 527, (2) 163-168 (2004) [3] Modolo, G.; Asp, H.; Schreinemachers, C.; Vijgen, H. Solv. Extr. Ion Exch., 25, (6) 703-721 (2007) [4] BABAIN, V.A.; ALYAPYSHEV, M.YU.; SMIRNOV, I.V.; SHADRIN, A.YU.. Radiochemistry, 48, (4) 369-373 (2006) [5] Paulenova, A.; Alyapyshev, M.Yu.; Babain, V.A.; Herbst, R.S.; Law, J.D. Sep. Sci. Technol., 43, (9) 2606-2618 (2008) [6] J. Rais, B. Grü Ion Exchange and Solvent Extraction, A Series of Advances, 17, 243-334, by Y. Marcus, A. ner, K. SenGupta, CRC Press, (2004). [7] Paulenova A., Alyapyshev, M. Yu, Babain, V. A. ·Herbst, R. S. ·Law, J. D. Solvent Extraction and Ion Exchange, Solvent Extraction and Ion Exchange, Volume 31, Issue 2, 184-197, 2013 21
  • Recovery of Uranium by Adsorbents with Amidoxime and Carboxyl Groups: A Density Functional Study Wei-Qun Shi*, Cong-Zhi Wang, Jian-Hui Lan, Zhi-Fang Chai Nuclear Energy Nano-Chemistry Group,CAS Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Institute of High Energy Physics, Beijing 100049, China * Corresponding author: shiwq@ihep.ac.cn In seawater, uranium is present mainly in the form of UO 2 (CO 3 ) 3 4- with a concentration of about 3-3.3 mg/L. Recovery of uranium from seawater has been studied over several decades [1, 2]. It has been found that adsorbents with amidoxime (HAO) groups show high tendency towards uranium, and the introduction of carboxyl (HAA) groups can increase the adsorption capacity of uranium. In this work, the adsorbent behavior of UO 2 2+ by adsorbents containing amidoxime and carboxyl groups have been studied by density functional theory (DFT) in conjunction with relativistic small-core pseudopotentials. Our results reveal that there are three binding modes for the amidoxime group, i.e. the monodentate coordination with the oxime oxygen atom, the bidentate coordination through the oxime oxygen and the amine nitrogen atoms, and th e η2 coordination via the N-O bond. As for the carboxyl group, it acts as monodentate and bidentate ligand to UO 2 2+. Additionally, amidoximes can form cyclic imide dioximes, which coordinate to UO 2 2+ as tridentate ligands. Natural bond orbital analysis and electron localization function analyses indicate that in these complexes there exist strong U-O and U-N bonding and the species with η2 coordination mode exhibit higher covalent character. As reported in the literature, the co-existence of amidoxime and carboxyl groups can enhance the adsorbability of uranium. The 1:4 (metal:ligand) type complexes are found to be the most stable species with the 1:1 stoichiometry of amidoxime and carboxyl. In these complexes, the amidoxime ligands prefer to coordinate in η2 binding mode to UO 2 2+. Moreover, our calculations also show that these adsorbents have higher adsorbability for vanadium than uranium, which is in accordance with the experimental results. [1] R. Sellin and S. D. Alexandratos, Ind. Eng. Chem. Res. 52, 11792 (2013). [2] H. Egawa, N. Kabay, T. Shuto and A. Jyo, Ind. Eng. Chem. Res. 32, 709 (1993). Fig. 1. Optimized structures of the uranyl complexes with adsorbents containing amidoxime (HAO) and carboxyl (HAA). This work was supported by the National Natural Science Foundation of China (Grant Nos. 21101157, 21201166, 11105162, 21261140335) and the “Strategic Priority Research program” of the Chinese Academy of Sciences (Grant Nos. XDA030104). 22
  • Theoretical Studies on the Electronic Structure and Chemical Bonding of UX 5– (X = F, Cl) Complexes Jing Sua,b*, Phuong Diem Dauc, Xiao-Gen Xionga,b, Lai-Sheng Wangc, Jun Lib* a Division of Nuclear Materials Science and Engineering, Shanghai Institute of Applied Physics,Chinese Academy of Sciences, Shanghai 201800, China b Department of Chemistry, Tsinghua University, Beijing 100084, China c Department of Chemistry, Brown University, Providence RI 02912, USA * Corresponding author: sujing@sinap.ac.cn (J.S.); junli@mail.tsinghua.edu.cn (J.L.). Molten salts are important in the nuclear energy industry both as coolants and in hydrometallurgical liquid-liquid extraction for reprocessing of spent fuels.[1]Fluoride- and chloride-based salts, such as LiF-BeF 2 andLiCl-KCl melts, are used in pyrochemical nuclear applications due to their radiolytic stability.[2,3] Knowledges of the electronic structures, and chemical and thermodynamic properties of uranium halides, especially fluorides and chlorides, are important to understanding the actinide chemical speciation and redox processes in molten salts. Here we report the gas-phase investigation of the electronic structures of UX 5 –(X = F, Cl) using photoelectron spectroscopy (PES) and relativistic quantum chemistry. [4,5]Theoretical investigations reveal that the ground states of UX 5 –(X = F, Cl) have an open shell with two unpaired electrons occupying two primarily 5f xyz andU 5f z 3based molecular orbitals (8a 1 and 2b 2 respectively, see Fig. 1). The structures of UX 5 – and UX 5 (X = F, Cl)are theoretically optimized and confirmed to have C 4v symmetry.The UX 5 – anionsare highly electronically stable with adiabatic electron binding energies of 3.82±0.05 eV and4.76±0.03eVfor X= F and Cl, respectively. An extensive vibrational progression from U-F symmetrical stretching mode isobserved in the spectra of UF 5 –, which is well reproduced by Franck-Condon simulation.Systematic chemical bonding analysesare performed on all the uranium pentahalide complexes UX 5 – (X= F, Cl, Br, I).The results indicate that the U-X interactions in UX 5 – are dominated by ionic bonding, with increasing covalent contributions for the heavier halogen complexes. Fig. 1.The two singly occupied molecular orbitals in UX 5 –(X = F, Cl) References [1] C. Le Brun, J. Nucl. Mater. 360, 1 (2007). [2] Y.H. Cho, T.J. Kim, S.E. Bae, Y.J. Park, H.J. Ahn and K. Song, Microchem. J. 96, 344 (2010). [3] M. Salanne, C. Simon, P. Turq, R.J. Heaton, P.A. Madden, J. Phys. Chem. B 110, 11461 (2006) [4] P.D. Dau, J. Su, H.T. Liu, D. L. Huang, F. Wei, J. Li and L.S. Wang, J. Chem. Phys. 136, 194304 (2012) [5]J. Su, P.D. Dau, C.F. Xu, D.L. Huang, H.T. Liu, F. Wei, L.S. Wang and J. Li. Chem. Asian J. 8, 2489 (2013). 23
  • First-principles calculation of intrinsic and defective properties of UO2 and ThO2 Han Hana, Cheng Chenga and Ping Huaia* a Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai, China 201800 *Corresponding author: huaiping@sinap.ac.cn The coated particle fuel is originally designed for the high temperature gas-cooled reactor. Recently, several new high temperature reactor concepts have been developed. For instance, small modular Advanced High Temperature Reactor is a new small modular fluoride salt cooled reactor concept developed at Oak Ridge National Laboratory. The Thorium Molten Salt Reactor (TMSR) in China has also proposed a concept design based on pebblebed fluoride salt cooled reactor with thorium-uranium alternate once-through fuel cycle. In the history of coated-particle fuel, the Tristructural Isotropic (TRISO) fuel is one of the most reliable candidates, which has a uranium oxycarbide kernel coated with a series of layers that act as the cladding. The inner pyro-carbon layer is designed to accept gaseous fission products and attenuate fission product recoils. The SiC layer’s function is to contain metallic fission products and provide structural support for the fuel particle. The outer pyro-carbon layer serves as a structural component and protects the SiC layer during compacting. These coated particle fuels have highly robust safety characteristics, with the ability to retain fission products up to temperatures of 1600°C or more. Uranium, thorium and plutonium fuels have been experimentally used in form of oxides, carbides and nitrides in TRISO particles. The behaviour of nuclear fuel in reactor is very complicated due to their neutronics properties as well as thermo mechanical strength, chemical stability, microstructure, and defects. It is very important to understand these material properties from microscopic picture. The complicated bonding nature of 5f-orbital leads to unique electronic structure of actinide compounds [1-2]. In this paper, the properties of intrinsic/defective uranium and thorium dioxide are studied by using the density functional theory in the generalized gradient approximation. A small lattice distortion is found due to the magnetic ordering of ground state of UO2 (as illustrated in Figure 1(a). The lattice constant c0 (parallel to the spin) is different from the other two constants a0 and b0. Strong correlation also plays an important role in UO2. The Hubbard U correction method has been introduced to describe the correlation. By taking into account the Hubbard U correction, the lattice constants are increased to a0=5.57 Å, and c0=5.50 Å. We have also checked the total phonon density of states in case of the small lattice distortion, which was obtained by minimizing the total energy of the electronic structure calculations. The dispersion curves of the distorted UO2 crystal has been shown in Fig. 1(b) with the LO-TO splitting. The U-dependence of the phonon density of states is found to be very weak, which is consistent with the theoretical assumption that excited-state properties of the electronic states should not affect ground-state materials properties very much. Figure 1 (a) The structure and magnetic ordering of ground state of UO2. (b) The phonon dispersion curves of antimagnetic UO2. The LO-TO splitting effect and Hubbard U correction are all concerned. References: [1] G. Schreckenbach, G. Shamov, Acc. Chem. Res. 43, 19 (2010). [2] Kevin T. Moore, Rev. Mol. Phys. 81, 235-298 (2009). 24
  • Modeling the autocatalytic reaction between TcO 4 - and MMH in HNO 3 solution Fang LIU, Hui WANG, Yan WEI, Yong-fen JIA (China Institute of Atomic Energy, Beijing, 102413) liuxinyu741@sohu.com Abstract An advanced PUREX process was innovated by China Institute of Atomic Energy, which adopts N, N-dimethylhyldroxylamine (DMHAN) as reducing agent and methyl-hydrazine (MMH) as stabilizer in U/Pu splitting stage. MMH is a moderate reductant, it may impact technetium valence so as to decide the technetium distribution in the process. This paper aimed at (i) investigating the reaction between TcO 4 - and MMH in HNO 3 solution with an autocatalytic reaction model. Two equations widely used for modeling autocatalytic reaction are adopted to simulate the reaction, and (ii) studying the concentration effects on kinetic parameters such as maximum reaction rate, lag time. Isothermal experiments were conducted at temperatures ranging from 25°C to 55°C using reactant solutions range wider than the real technology. This concentration effect was included in the proposed kinetic models, which were able to successfully describe experimental data, and further more may be able to predict technetium behaviour in U/Pu splitting stage. The advanced salt-free PUREX process adopts MMH as stabilizer, one of the reason is to avoid producing hydragoic acid which is a product of Hydrazine oxidation. Prior works proved that DMHAN can not reduce Tc(VII) in HNO 3 solution [1], and in the U/Pu splitting stage of advanced salt-free PUREX technetium goes into aqueous solution mainly in Tc(IV) form. The reaction between technetium and hydrazine has been studied by many investigators [2, 3], so we presume that Tc(VII) was mainly reduced by MMH in this system. In this paper the reaction between technetium and MMH was studied detailed in the aspect of Tc(VII) concentration. A typical c-t curve of TcO 4 - reduced by MMH in HNO 3 solution is presented in figure 1. The X axes stands for the growth of low valence Tc. 25000 20000 count(cpm) 15000 10000 experiment dates logic fit gompertz fit 5000 0 2 0 2 4 6 8 10 12 14 time (hour) Fig 1. Growth of low valence technetium depend on time 40°C, c 0 (Tc(VII))=7.4×10-4mol/L, c 0 (HNO 3 )=1.5mol/L, c 0 (MMH)=0.15mol/L This is a typical S mode curve in the reaction of MMH deoxidize Tc(VII). There is a leg phase in the initial moment, then the Tc(VII) concentration declined sharply, in the end of the reaction Tc(VII) concentration decreases slowly again. The logic equation and Gompertz function 25
  • are widely used for simulating a sigmoid curve. In figure 1 both the two equations can simulate the experiment dates very well. The R-Square is 0.9972 for logic simulation and 0.9951 for Gompertz simulation. The logic equation, Y=a/(1+exp(k*(x-x 0 ))), is one solution of equation dy/dx = k*y*(c 0 -y). And P. D. Willson used equation dTc(VII)/dt = k*Tc(VII)*Tc(IV) to simulate the reaction between TcO 4 - and MMH in HNO 3 solution. The experiment dates can be simulated by the logic equation, it indicates that the reaction of MMH reducing TcO 4 - may be a autocatalytic mode. The low valence Tc, mainly Tc(IV) act a important role in this reaction. The influence of initial Tc concentration, MMH concentration and acidity on the reaction is studied in this paper. The initial Tc concentration has a distinct effect on parameter k of both equations, and has slight effect on x 0 . MMH concentration and acidity effect the x 0 remarkable than k. A higher MMH concentration will get a little x 0 value. There is a proper acidity in this reaction, in this acidity there is a smallest x 0 value. For the mathematics aspect, the parameter k mainly charges the maximum of reaction velocity. The parameter x 0 mainly charges the log period time. That is to say, the initial Tc concentration mainly affects the maximum of reaction velocity, and the MMH concentration and acidity decide when the fastest reaction happens. The influence of MMH concentration is showed in figure 2 and table 1. 50000 40000 count (cpm) 30000 1 2 3 4 5 20000 10000 0 0 2 4 6 8 10 time (hour) Fig.2 The influence of MMH concentration on reaction 40°C, c 0 (Tc(VII))=7.4×10-4mol/L, c 0 (HNO 3 )=1.5mol/L, c 0 (MMH): 1--0.068M, 2--0.15M,3--0.225M, 4--0.34M, 5--0.51M Table 1. The influence of MMH concentration on the parameter of simulation logic simulation Function: y=a/(1+exp(k*(x-x 0 ))) C MMH a k R2 x0 43297 43831 0.225M 46478 0.51M 45318 0.068M 0.15M 0.90 0.84 0.85 0.97 Gompertz simulation Function: y=a*exp(-exp(-k*(x-x 0 ))) a k R2 x0 5.26 0.9974 4.36 0.9975 3.77 0.9978 3.01 0.9930 47290 0.51 4.69 0.9927 50036 0.45 3.87 0.9951 51627 0.48 3.22 0.9964 49068 0.58 2.49 0.9977 References [1] Fang LIU, Master's Thesis of CIAE, 2009 [2] P. D. Wilson, J. Garraway, in Proc. Int. Meet. On Fuel Reprocessing and Waste Management, La Grange Park, (the United States), 1984, vol. 1, p. 467. [3] J. Garraway, Journal of the Less Common Metals, Volume 97, February 1984, pp. 191–203. 26
  • Fluorescent BINOL-Based Sensor for Thorium Recognition and a Density Functional Theory Investigation Jun Wen, Liang Dong, Sheng Hu, Tong-Zai Yang , Xiao-Lin Wang * Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang, 621900, Sichuan Province, China * Corresponding author: xlwang@caep.ac.cn. Because of the widespread use of thorium and its toxic properties, the development and improvement in analysis methods for the determination of thorium would be useful [1-3]. We developed a novel 1,1′-bi-2-naphthol (BINOL) derivative fluorescence sensor L-1 for the recognition of thorium ion with a fluorescence quench response. This ligand showed high selectivity and sensitivity for thorium ion recognition (Figure 1). When an equivalent of Th4+ was added to the solution of L-1, dramatic fluorescence quenching ( quenching efficiency: 64%) was observed, suggesting that compound L-1 showed a specific response with Th4+ ions due to the chelation-enhanced fluorescence quenching (CHEQ) effect. This is the first one-to-one stoichiometric responding chemical sensor for thorium, and indicated a 1:1 bonding mode between L-1 and Th4+ ions (Figure 2). The detection limit [4] of L-1 for the determination of Th4+ was estimated to be 6 × 10−7 M in 1:1 MeOH:H2O (v/v). Moreover, the binding constant (K) derived from the fluorescence titration data was found to be 3.4 × 103 using a Benesi–Hildebrand plot [5]. According to previous reports [2], many thorium sensors have encountered interference by uranyl ions. Nevertheless, L-1 displayed good selectivity for thorium. To further understand the nature of the binding interactions of Th4+ and UO 2 2+ with the ligand, coordination effects were investigated by DFT calculations. According to these analyses of structures, electronic properties, and energetics, we can conclude that the binding interaction between L-1 with Th4+ is stronger than that with UO 2 2+, and that the L-1 ligand forms a stable complex with Th4+. [1] Handbook of Hazardous Materials (Ed: M. D. Corn), Academic Press, San Diego 1993. [2] A. Safavi, M. Sadeghi, Anal. Chim. Acta, 567, 184-188, (2006). [3] F. S. Rojas, C. B. Ojeda, Anal. Chim. Acta, 635, 22-44, (2009). [4] V. Thomsen, D. Schatzlein, D. Mercuro, 18, 112-114, (2003). [5] H. A. Benesi,; J. H. Hildebrand, J. Am. Chem. Soc. 71, 2703-2707, (1949). 27
  • Exceptional Selectivity for Actinides by N,N’-Diethyl-N,N’-Ditolyl-2,9-Diamide-1,10Phenanthroline Ligand: A Combined Hard-Soft Atoms Principle# Cheng-Liang Xiao, Li-Yong Yuan, Yu-Liang Zhao, Zhi-Fang Chai, Wei-Qun Shi* Key Laboratory of Nuclear Radiation and Nuclear Energy Technology and Key Laboratory For Biomedical Effects of Nanomaterials and Nanosafety, Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049, China * Corresponding author: shiwq@ihep.ac.cn MA(III) and Ln(III) have similar physicochemical properties, such as oxidation state, ionic radii, hydration, and complexation mode. Extractants containing soft sulfur or nitrogen atoms are preferred to recognize MA(III) over Ln(III). R-BTP, R-BTBP, and R-BTPhen ligands are the successful representatives for Ln(III)/MA(III) in the last 20 years [1-2]. However, light actinides (U, Np, Pu) normally favor ligands (alkylamide or alkylphosphate) containing hard oxygen atoms [3]. If we make sure the selectivity for light actinides using hard-atoms ligands, the separation of MA(III) from Ln(III) is difficult to achieve. To separate all the actinides from lanthanides, the synthesis, solvent extraction, and complexation behaviors of actinides and lanthanides by a novel phenanthroline-based tetradentate ligand with combined hard-soft atoms, N,N’-diethyl-N,N’-ditolyl-2,9-diamide-1,10-phenanthroline (Et-Tol-DAPhen, 1), are described in this work. The ligand exhibits excellent extraction ability and high selectivity of actinides over lanthanides from highly acidic solution. X-ray crystallographic structures of Et-Tol-DAPhen with thorium and uranyl ions are showed to be 1:1 complexation. The stability constants for some actinides and lanthanides complexes with Et-Tol-DAPhen are also determined in methanol by UV-Vis spectrometry. The results of density functional theory (DFT) calculation (Fig. 1) reveal that the An-N bonds have more covalent characters than that of Eu-N, which may dominate the selectivity of Et-Tol-DAPhen towards actinides. This work can shed light on the design of new ligands with combined soft-hard atoms for group separation of actinides from highly acidic nuclear waste. [1] J. H. Lan, W. Q. Shi, Z. F. Chai , et al., Coord. Chem. Rev., 256, 1406 (2012). [2] M. J. Hudson, L. M. Harwood, D. M. Laventine, et al., Inorg. Chem., 52, 3414 (2013). [3] C. Z. Wang, J. H. Lan, W. Q. Shi, et al., Inorg. Chem., 52, 196 (2013). (a) (b) (c) (d) Fig. 1. Optimized structures of (a) Am(1)(NO 3 ) 3 , (b) Eu(1)(NO 3 ) 3 , (c) [UO 2 (1)(NO 3 )]+, (d) Th(1) -(NO 3 ) 4 by B3LYP/6-311G(d,p)/RECP method in gas phase. # This work was supported by NSFC (Grants 91026007, 21201166, and 21101157) and the "Strategic Priority Research Program" of the Chinese Academy of Sciences (Grant XDA030104). 28
  • The studies on optimization of the separation method of Am and Cm Zhuoxin Yin, Ping Li, Wangsuo Wu* Radiochemistry Laboratory, Lanzhou University, Lanzhou 730000, Gansu, China *Corresponding author: wuws@lzu.edu.cn With the development of the nuclear industries, the amount of the spent nuclear fuel continued to grow. According to the concept of advanced nuclear fuel cycle, every element of the HLLW should be separated independently. The method of the element separation depended upon the chemical species of the element and the composition of the samples. Americium and curium are two kinds of highly radioactive and long half-life elements in the HLLW. Because of the similar chemical properties of americium and curium, they are hardly to be segregated. Many techniques had been taken to solve the problem of the separation of Am and Cm, each of them had its advantages and disadvantages. The Solvent Extraction[1] was the most common way for separating Am and Cm, sometimes it used batch and column methods, the lately investigation aimed to synthesize the new ligands and find new complex structures in order to get better spatial results than before. The ionic liquids[2], electrodeposition[3] or adsorption were also taken part in laboratory experiments, in order to find new ways to separate Americium and Curium. Generally, the existences of most transuranium elements in the HLLW were trivalent ions. The Valence Control[4] in some situation might be result in good separated effect. For the sake of the optimization of the separation method of Am and Cm, further research should be made and more experimental data should be obtained. 4.50x10-6 4.50x10-6 4.00x10-6 4.00x10-6 3.50x10-6 3.50x10-6 3.00x10-6 3+ Am Am(OH)+ 2 Am(OH)3 2.50x10-6 2.00x10-6 Species(mol/L) Species(mol/L) 3.00x10-6 Am(OH)2+ 1.50x10-6 1.00x10-6 Cm3+ Cm(OH)+ 2 2.50x10-6 Cm(OH)2+ 2.00x10-6 1.50x10-6 1.00x10-6 5.00x10-7 5.00x10-7 0.00 0.00 -5.00x10-7 0 1 2 3 4 5 6 7 8 9 10 11 12 13 0 14 1 2 3 4 5 6 7 8 9 10 11 12 13 pH pH (a) Species of Am(III) (b) Species of Cm(III) Fig.1. Speciation distribution of Am(III) as function of pH in aqueous solution References: [1] Y. Sasaki, Y, Kitatsuji, Y. Tsubata, Y. Sugo, Y. Morita, Solvent. Extr. Res. Dsv. 18, 93(2011). [2] K. Binnemans, Chem. Rev, 107, 2592(2007). [3] S. Liu, Atomic. Ene. Sci. Technol. 22, 238(1988) [4] K. Marmoru, F. Tetso, K. Fumio, J. Nucl. Sci. Technol. 35, 185(1998). 29 14
  • Burn-up calculation of plutonium in fusion-fission hybrid reactor Kento Fukanoa*, Shunji Tsuji-Iioa, Hiroaki Tsutsuia, Yoji Someyab a Tokyo Institute of Technology b Japan Atomic Energy Agency *fukano.k.aa@m.titech.ac.jp Introduction Nuclear power plants (NPPs) are important as base load power source in the world even after Fukushima Daiichi nuclear disaster. But NPPs have a problem in terms of nuclear proliferation because plutonium (Pu) is produced when NPP generates electric power. Therefore, Pu should be burned up by some method. On the other hand, ITER which is the international fusion experimental reactor uses 10 kg of tritium (T) as fuel in the start-up phase, and it will exhaust 21 kg of T existing in the world. The world is deficient in T. Therefore T should be produced by some method. A solution to solve these two problems is fusion-fission hybrid reactor. Hybrid reactor can burn up Pu and breed T effectively. The objective of this study is to put forward scenarios for Pu burn-up in terms of nuclear non-proliferation and fusion reactor introduction from the aspect of tritium supply in the world. As preparatory, this study designs hybrid reactors for Pu burn-up and T production with simulation codes. Assumed type of fusion reactor is tokamak by magnetic confinement. The blanket in fusion reactor in this study is comprised of MOX fuel, Li2TiO3, water, F82H, and SUS316LN. Design requirements There are four requirements to design feasible hybrid reactors for Pu burn-up and T breeding, the life time of magnetic field coil is over 40 years, tritium breeding ratio (TBR) is over unity, and the amount of burn-up plutonium per year is over 7 t, the temperature of each composition material is below its upper temperature limit. The first target of this study is to find out hybrid reactor parameters to meet the above requirements with simulation codes. Result and consideration According to simulation results, the life time of magnetic field coil is 50 years, TBR is 4.08. Figure 1 shows a reduction in the total amount of plutonium. The initial loading plutonium is 22 t. After 1 year, it is reduced to 13 t , so that the amount of burn-up plutonium per year is 9 t. Figure 2 indicates the temperature distribution of MOX fuel layers in blanket. The layers include MOX fuel and water in alternate. The upper temperature limit of MOX fuel is 2500℃. Therefore this blanket is feasible thermally as well as other layers. This designed hybrid reactor fulfills the four requirements. The next step is to make up operation scenarios of this hybrid reactor. Fig. 1. Time evolution of total Pu amount Fig. 2. Temperature distribution in blanket 30
  • [UO2(NO3)4]2- Complex in Ionic Liquids Investigated by Optical Spectroscopic and Electrochemical Studies Yupeng Liu, Taiwei Chu* Beijing National Laboratory for Molecular Sciences, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, China. * Corresponding author: twchu@pku.edu.cn. The tetranitratouranium(VI) complex, [UO2(NO3)4]2-, is thought to be unstable in common molecular solvents. Recently, C. Gaillard et al. [1] has proven the formation of [UO2(NO3)4]2- in the ionic liquid [BMI][NO3] by EXAFS study. However, the character of this complex in ILs is still unknown. In this report, we studied the optical spectra of [UO2(NO3)4]2-, calculated the formation constants in various ILs, and investigated its electrochemical behaviors. The UV-vis spectrum of [UO2(NO3)4]2- (Fig.1) shows a strong ‘continuous’ broad band in the 380~480 nm region, far from the remarkable sharp vibronic bands of [UO2(NO3)3]- [2]. It can also formed in hydrophobic ILs ([NTf2]-- and [PF6]--based) with excess of [NO3]-. The luminescence of [UO2(NO3)4]2- in non-imidazolium ILs is much stronger than that of [UO2(NO3)3]-, and without vibronic fine structures. By quantitative analysis of UV-vis spectra of serial samples with various nitrate concentrations, the equilibrium constant (K4) of [UO2(NO3)3]- + [NO3]- = [UO2(NO3)4]2- can be gotten. The constants in several hydrophobic ILs are ranging from 10 to 30 (Table.1), much higher than in molecular solvents. ILs with aromatic cations show lower K4 values, because these cations have stronger interactions with the planar [NO3]- anion. Table 1. Formation constant values of [UO2(NO3)4]2- in several [NTf2]—based ILs. Cations Aromatic K4 BMI Y 10.34 BDMI Y 12.08 BPy Y 12.79 N4111 N 15.35 N4221 N 20.04 Pyr14 N 26.48 PP14 N 30.12 CH3NO2 * 4.74 * Nitromethane as solvent, from Ref.[3]. Fig.1. UV-vis spectra of [UO2(NO3)4]2- in [BMI][NO3] and [UO2(NO3)3]- in [BMI][NTf2]. Vibrational spectra show more detailed information on the interactions between [NO3]- and uranyl. The notable redshift (8~10 cm-1 vs. [UO2(NO3)3]-) of uranyl stretching frequencies in both symmetry (Raman) and asymmetry (infrared) modes (Fig.2) indicates the stronger interaction from equatorial ligands. Moreover, the ATR-FTIR spectrum in [NO3]- stretching region shows the presence of two kinds of coordinated [NO3]- in [UO2(NO3)4]2-. In either [BMI][NO3] or [Pyr14][NO3]/[NTf2] (1.2M [NO3]-), [UO2(NO3)4]2- shows a quasi-reversible U(VI)/U(V) electrochemical redox process. In [Pyr14][NO3]/[NTf2], the half-wave potential of U(VI)/U(V) is -1.10 V (vs. Ag+/Ag, 308K), the Ipc/Ipa ratio is ~0.7 while scan rate varies from 0.01 to 0.10 V/s (Fig.3 and Table 2), and the diffusion coefficient D is (2.10±0.06)×10-8 cm2/s. 31
  • The notable stability of [UO2(NO3)4]2- in ionic liquids suggests that this complex may play an important role in the NFC processes involving ILs containing nitrate anion, and this complex may have potential in the development of IL-based electrochemical separation and purification processes. Fig.2. ATR-FTIR (left) and Raman (right) spectra of [UO2(NO3)4]2- (black) and [UO2(NO3)3](grey) in the O=U=O stretching region. Table 2. Reversibility of U(VI)/U(V) redox of [UO2(NO3)4]2- in[Pyr14][NO3]/[NTf2]. Scan rate Ε1/2 ∆Ep Ipa/Ipc V/s V* mV 0.01 -1.105 154 0.65 0.02 -1.100 146 0.68 0.03 -1.098 143 0.70 0.05 -1.096 137 0.70 0.07 -1.095 132 0.69 0.10 -1.095 135 0.67 + * Potential against Ag /Ag. Fig.3. Cyclic voltammograms of [UO2(NO3)4]2at various scan rates in [Pyr14][NO3]/[NTf2]. Insert: the linear relationship of Ipc against square root of scan rate. T = 308K. [1] C. Gaillard, O. Klimchuk, A. Quadi, I. Billard and C. Hennig. Dalton Trans., 41, 5476 (2012) [2] K. Servaes, C. Hennig, I. Billard, C. Gaillard, K. Binnemans, C. Gorller-Walrand, and R. Van Deun. Eur. J. Inorg. Chem., 2007, 5120 (2007) [3] J. L. Ryan. J. Phys. Chem., 65, 1099 (1961) 32
  • Complexation of Uranyl by Neutral Bidentate Phosphonate Ligands in Ionic Liquids Yupeng Liu, Taiwei Chu* Beijing National Laboratory for Molecular Sciences, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, China. * Corresponding author: twchu@pku.edu.cn. The potentiality of ionic liquids (ILs) in the nuclear industry has been explored in recent years, especially for the extraction of uranium from aqueous medium using ILs [1]. Knowledge on the interactions between uranyl and the extracting agents (ligands) in ILs is important to understand the extraction progress. Recently, we have reported a unique 2:1 dicationic complex, [UO 2 (TEMBP) 2 ]2+, formed by a bidentate ligand tetraethyl methylenebisphosphonate (TEMBP) and uranyl in ILs [2].The optical spectra and electrochemistry of uranyl complexes of some monodentate organophosphorus ligands have also been studied by our group [3]. In this report, we studied the uranyl complexes formed in [BMI][NTf 2 ] with bidentate ligands related to TEMBP, the compete complexation between chelate ligands and NO 3 , and the spectra of uranyl complexes extracted from nitrate solutions. Fig.1 shows the UV-vis spectra of uranyl complexes formed from UO 2 (ClO 4 ) 2 . Complexes similar with [UO 2 (TEMBP) 2 ]2+ are formed by bidentate liands such as tetrabutyl methylenebisphosphonate (TBMBP) and tetrabutyl ethylenebisphosphonate (TBEBP). Since these complexes have similar structure, their spectra (b, c, d) resemble each other. Their spectra also have some similarity with those of [UO 2 (TBP) 4 ]2+ and [UO 2 (DBBP) 4 ]2+ (a, b), because they all have tetragonal coordination to the uranyl by P=O groups [2,3]. Fig. 1. UV-vis spectra of uranyl complexes in [BMI][NTf 2 ]. (a), [UO 2 (TBP) 4 ]2+; (b), [UO 2 (DBBP) 4 ]2+; (c), [UO 2 (TEMBP) 2 ]2+; (d), [UO 2 (TBMBP) 2 ]2+; (e), [UO 2 (TBEBP) 2 ]2+. Fig. 2. UV-vis spectra of samples in [BMI] -[NTf 2 ]. (a) ~ (e), uranyl nitrate with 1 to 10 eq. of TEMBP; (f), [UO 2 (TEMBP) 2 ]2+; (g), [UO 2 (TBP) 4 ]2+; (h), uranyl nitrate with 1M TBP; (i), [UO 2 (NO 3 ) 2 (TBP) 2 ] in pure TBP. Results of compete complexation between ligands and NO 3 - are showing in Fig.2. With 1 equivalent molar of TEMBP added, [UO 2 (NO 3 ) 2 (TEMBP)] complex is formed in [BMI][NTf 2 ] (a). The spectra then change gradually with increasing TEMBP concentrations, as evidenced by the shrink of characteristic bands of [UO 2 (NO 3 ) 2 (TEMBP)] (b ~ e) and emerging of new bands those belonging to [UO 2(TEMBP) 2 ]2+ (f). The changes in spectra indicate that excess of TEMBP can replace the coordinated NO 3 - in [UO 2 (NO 3 ) 2 (TEMBP)] to form [UO 2 (TEMBP) 2 ]2+ in the IL. In contrast, [UO 2 (NO 3 ) 2 (TBP) 2 ] is the complex formed by uranyl nitrate with even large excess of TBP (h), with its spectrum similar with that of [UO 2 (NO 3 ) 2 (TBP) 2 ] in pure TBP (i) and much different from [UO 2 (TBP) 4 ]2+ (g). TBMBP and 33
  • TBEBP also show similar substitution ability. Information on the interaction between ligands and uranyl can be obtained by IR spectra (Fig.3). In [UO 2 (NO 3 ) 2 (TEMBP)], the P=O group is coordinated to the uranyl thus its stretching band shifts to lower wavenumbers. With excess of TEMBP added, the band due to free P=O group (1258 cm-1) -1 appears. The ligand substitution is evidenced by the decrease of intensity of band at 1525 cm , which is the υ(NO) stretching band of coordinated NO 3 - (in bidentate mode) [4]. In the case of TBP as ligand, the υ(NO) band almost does not change with increasing TBP concentration. Fig.4 UV-vis spectra of [UO 2 (TBMBP) 2 ]2+ (a) and uranyl extracted by 0.1M TBMBP /[BMI][NTf 2 ]. Aqueous HNO 3 solutions are (b) = 0.01M, (c) = 1M, and (d) = 6M. Initial uranyl concentration in the aqueous solutions is 0.01M. Fig.3. ATR-FTIR spectra of uranyl nitrate with 1, 3 and 10 eq. of TEMBP in [BMI][NTf 2 ] in the υ(P=O) (left) and υ(NO) region (right). The insert graph shows spectra of uranyl nitrate with various amount of TBP in the υ(NO) region. Unlike the TBP/IL system, extraction of uranyl by TBMBP/[BMI][NTf 2 ] is less dependent on the aqueous HNO 3 concentration. With 0.1M TBMBP/[BMI][NTf 2 ], almost 100% of the uranyl is extracted, while the acid concentrations ranging from 0.01M to 6M. Spectra of the IL phase after extraction are showing in Fig.4, and are all similar with the spectrum of [UO 2 (TBMBP) 2 ]2+, suggesting that extraction of uranyl by TBMBP via a single mechanism independent on HNO 3 concentration. The ability to substitute coordinated NO 3 , and the HNO 3 -independent extraction mechanism, indicate that neutral bidentate ligands have enhanced coordinating ability to uranyl versus their monodentate analogs such as TBP. The high efficiency and single mechanism of extraction in ionic liquids make them potential better alternatives for TBP. [1] I. Billard, A. Quadi, and C. Gaillard. Anal. Bioanal. Chem., 400, 1555 (2011). [2] Y. Liu, T. Chu, and X. Wang. Inorg. Chem., 52, 848 (2013). [3] Y. Wang . Studies on the Optical Spectra and Electrochemistry of Uranyl Complexes in Ionic Liquids. Master’s Thesis, Peking Universtiy, 2013. [4] K. Nakamoto. Infrared and Raman Spectra of Inorganic and Coordination Compounds, Theory and Applications in Inorganic Chemistry. Wiley-Interscience. 2009 34
  • Session 3: Waste Management 35
  • Sorption of Uranium and Rhenium in the Presence of Fulvic Acids Troshkina I.D., Shilyaev A.V., Grechov A.P. D. Mendeleyev University of Chemical Technology of Russia, Miusskaya Sq. 9, Moscow 125047, Russia. E-mail: tid@rctu.ru Nuclear fuel cycle starts with the recovery and processing of uranium ores. Implementation of the first stage of the nuclear fuel cycle (NFC) - uranium mining is accompanied by a strengthening of the role of complex processing. Rhenium, an important less-common metal, has a special place among the other by-product metals due to the high profitability of its recovery [1,2]. Intensification of leaching method is possible by the introduction of a natural nutrient solutions of fulvic acids (FA) [3], which is accompanied by a savings leaching agent - sulfuric acid, improving environmental conditions of the process. The degree of recovery of uranium ore into productive solutions increases. The information on the influence of the fulvic acids on the subsequent sorption recovery of uranium are limited and rhenium - none. The aim of this presentation - to assess the impact of fulvic acids on the equilibrium sorption properties of strongly basic ion-exchange resins in the recovery of rhenium from uranium-containing mineralized solutions. Sorption of uranium and rhenium from mineralized solutions by the strongly ionites Rossion 62 (Russia), Purolite A-600 (United Kingdom) and Lewatit K 6367 (Germany) in the presence of natural fulvic acids was studied. Ion exchangers Purolite A- 600 and Lewatit K 6367 have a gel structure and belong to the type I (trimethylbenzylammonium functional groups). The ion exchanger Rossion - 62 has gel structure, the functional groups – pyrindinium. Sorption of metals by strongly basic ion exchangers of different types from mineralized solution containing sulfate ions - 10 g/l, chloride ion - 1 g/l was studied under static conditions at room temperature. Acidity of solution adequate pH 3. The concentration of fulvic acid in the solution was varied from 25 to 100 mg/l. To evaluate the effectiveness of sorption of metals from the mineralized solutions in the presence of fulvic acids was proposed the value  (%), calculated by the following formula:  = 100 - (E / E ) × 100 , where E - sorption capacity of the resin on metal, mg/g saturated from solutions without FA; E change of sorption capacity equal to the difference of the ion exchanger capacity in solution without FA and capacity in the presence of FA, mg/g. The dependence of the calculated values of the efficiency of sorption of uranium strongly basic ion exchangers on the concentration of FA in mineralized solution is shown in Fig. 1. It was found that the uranium capacity of ion exchangers is decreased with increasing concentration of FA. The least influence of FA was observed by the sorption of uranium by ion exchanger Rossion 62. Apparently, this contributes to the specific structure of the ion exchanger with a narrow pore distribution. Sorption of rhenium in the presence of FA by selected ion exchangers is unchanged. Fig. 1. The dependence of the efficiency of sorption of uranium from mineralized solutions (pH 3 ) on the concentration of FA. References [1]. A.A. Palant, I.D. Troshkina, A.M.Chekmarev, Metallurgy of Rhenium, (Science, Moscow, 2007). 298 p. (rus.) [2] In-Situ Leaching of Ores, (Academy of Mining Sciences, Publishing House, Moscow, 1998). 446 p. (rus.) [3] V.M. Panteleev. Thesis of IV Intern. conf. « Аtomeco-2010» 2010, Мoscow. P. 37. , 36
  • Oxalic acid effect on the diffusion of Se(IV) and Re(VII) in bentonite Tao Wu a*, Yong Luo a,b, Hai Wang a,b, Qing Zheng a, Yao Lin Zhao b , Jin Ying Li c a Department of Chemistry, Huzhou Teachers College, Huzhou 313000, P. R. China b School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049, P. R. China c China Resources New Energy Group Co., Ltd., HongKong, China * Corresponding author: twu@hutc.zj.cn Selenium-79 (6.5 × 10-5 years) and Technetium (2.1 × 105 years) are the concerned fission products in nuclear waste repositories. Se(IV) and Tc(VII) are the dominant species under enviromental condition, respectively. They have highly potential migration capacity from repository to geosphere because of the anionic exclusion between their anionic species and the nagative surface of bentonite [1]. Humic substances occur ubiquitously in soil and aquatic environments. They play a major role in controlling the physical and chemical characteristics of radionuclides because they have a variety of functional groups, such as –OH or –COOH functional group, which could form active parts of the bentonite. Under high temperature, pressure and radiation conditions, humic substances might be decomposed to smaller molecules. Some of them retain –OH or –COOH ions [2]. In basis to understand how light organic molecules can affect the diffusion behavior of Se(IV) and Re(VII) (analogue to Tc(VII)). Oxalate was used in this study. The diffusion behavior of Se(IV) and Re(VII) were investigated in present of oxalate at 1300 kg/m3 dry density by through diffusion methods under air conditions at pH around 3.5. Se(IV) and Re(VII) in source reservior were aged with oxalate in 0.1M NaClO 4 for 40 days. The sythnthetic pore water was used in target reservior. Figure 1 shows the flux (µg/cm2∙d) and the accumulated mass (Acum (µg)) as a function of diffusion time for Se(IV)-oxalate (Fig.1A) and Re(VII)-oxalate (Fig.1B) diffused in GMZ bentonite. Both the flux of Se(IV)-oxalate and Re(VII)-oxalate reached a peak at 8 days and 2 days, respectively. For Se(IV)-oxalate, Acum increased fast before the flux reached a peak. It slowly increased with increasing of diffusion time. The red precipitation was found in source reservior of Se(IV)-oxalate solution after diffused for several days. It could be explained that part of Se(IV) are retarded by the oxalate-carbonate-bentonite system. Whereas, Acum of Re(VII)-oxalate almost remained unchanged. No precipitation could be found in the source and target reserviors. It indicated that the diffusion of Re(VII)-oxalate were retarded in present of sythnthetic pore water. Re(VII) might sorbed on bentonite. Since carbonate was the dominant component in sythnthetic pore water, the Se(IV)-oxalate-carbonate and Re(VII)-oxalate- carbonate ternary compounds might formed. Therfore, Oxalate can retard the migration of Se(IV) and Re(VII) in GMZ bentonite in pore water. 37
  • Fig. 1 Flux and A cum vs. time for Se(IV)-oxalate and Re(VII)-oxalate at 1300 kg/m3 compacted GMZ bentonites by through-diffusion method. Acknowledgement This work was financially supported by the National Science Foundation for Distinguished Young Scholars of China (Grant No. 21207035). Reference 1. Wu, T., Wang, H., Zheng, Q., Zhao, Y. L., and Loon, L. R. V. Diffusion behavior of Se(IV) and Re(VII) in GMZ bentonite, Appl Clay Sci: submit 2. Institut fuer Nucleare Entsorgung (2000) Influence of humic acids on the migration behaviour of radioactive and non-radioactive substances under conditions close to nature: Metal-Ion behaviour in water/mineral system. Final Report 38
  • Migration of Actinides and fission products in Environments Toshihiko Ohnuki a AResearch Group for Bioactinides Chemistry, Advanced Science Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki, 319-1195, Japan * ohnuki.toshihiko@jaea.go.jp The presence of actinides (ANs) and fission products (FPs) in radioactive wastes is a major environmental concern due to their long radioactive half-lives, and its chemical toxicity. In order to estimate the mobility of ANs and FPs several studies examined its interactions with soils and subsoils composed of abiotic and biotic components, principally minerals and bacteria [1,2]. Among the biotic components, several microorganisms have been shown to sorb ANs and FPs[2]. The high capacity of microbial surfaces to bind ANs and FPs may affect the migration of ANs and FPs in the environment. Unfortunately, we have only limited knowledge of the role of microorganisms in the migration of ANs and FPs in the environment. We have studied the interaction of actinides and lanthanides with microorganisms for last two decades. We have elucidated that Bacillus subtilis poses higher performance on the adsorption of U(VI) and Pu(VI) in the mixture of B. subtilis and a clay mineral of kaolinite [3,4]. U(VI) is accumulated on the cells surafce of Saccharomyces cerevisiae to form H-autunite by the reaction with intracellular phophorous released by stimulation of U(VI) [5]. Cerium(III), analogue of Am and Cm, is adsorbed by Mn oxidizing bacteria. The adsorbed Ce is moved to MnO2 formed by the bacteria, and is oxidized to Ce(IV) [6]. These results show that microorganisms play important role for the retardation actinides in groundwater. After Fukushima Daiichi Nuclear Power Plant accident, migration of FPs should be clarified, especially radiocesium. So, many Japanese researchers have measured spatial distribution to obtain an initial mapping of radiocesiumin around the FDNPP [7-9]. The deposited radiocesium migrates in environment with time. However, the migration of radiocesium has not be fully understood. In the present study, chemical states of radioactive Cs, one of the FPs, in the contaminated soils by FDNPP accident has been characterized by the desorption experiments using appropriate reagents solutions and size fractionation of the contaminated soils [10]. More than 65% of radioactive Cs were remained in the residual fraction of the soil samples after treatment of a 1 mole L-1 NH 4 Cl solution and a 1 mole L-1 CH 3 COOH solution. Approximately 70% of radioactive Cs in the residual fraction were associated with the size fractions larger than the elutriated one, even though mica like minerals were contained in the elutriated one. These results strongly suggest that radioactive Cs was irreversibly associated with soil components other than mica like minerals in the contaminated soil. Subsequently, we have studied the adsorption behavior of radioactive Cs by the non-mica minerals kaolinite, halloysite, chlorite, montmorillonite, mordenite, MnO 2 , TiO 2 , Al 2 O 3 , and FeOOH to elucidate the environmental behavior of fallout radioactive Cs [11]. The adsorption and desorption experiments of Cs on the minerals were carried out at the Cs concentrations 1 x 10-4, 1 x 10-5 and 2 x 10-9 mole L-1 at pH 5.5. The desorption of Cs from the minerals was examined using 0.1 mole L-1 LiCl, NaCl, KCl, RbCl, and CsCl solutions. The sequential desorption was examined using a 0.1 mole L-1 LiCl solution, a 1 mole L-1 KCl solution, and a 1 mole L-1 HCl solution. The distribution coefficient (K d ) for the minerals at the Cs concentration 10-9 mole L-1 was in the order of mordenite > illite > montmorillonite, sericite, MnO 2 , kaolinite, and halloysite > chlorite, TiO 2 , Al 2 O 3 , and FeOOH, differing from the order observed at higher Cs concentrations. After the sequential desorption by the three reagent solutions, the residual fraction of Cs was lower at the Cs concentration 10-9 mole L-1 than at higher concentrations. Approximately 40%, 40%, 39
  • 50%, and 25% of the adsorbed Cs were residual in montmorillonite, mordenite, MnO 2 and kaolinite, respectively after the sequential desorption. These results strongly suggest that (1) radioactive Cs at 10-9 mole L-1 is more strongly associated with the non-mica minerals than at higher concentrations of 1 x 10-4 and 1 x 10-5 mole L-1, and (2) the non-mica minerals montmorillonite, mordenite, kaolinite, and MnO 2 contributed to the fixation of the radioactive cesium fall-out on Fukushima soil. [1] Waite, T.D., Davis J. A., Payne T. E., Waychunas G. A., Xu N. Geochim. Cosmochim. Acta. 58:5465-5478, 1994. [2] Suzuki Y. and Banfield J.F. Review in Mineralogy, 38. Washington DC, Mineralogical Society of America. pp. 393-432, 1999. [3] T. Ohnuki, T. Yoshida, T. Ozakia, M. Samadfam, N. Kozai, K. Yubuta, T. Mitsugashira, T. Kasama, A. J. Francis, Chem. Geol., 220, 237-243, 2005. [4] T. Ohnuki, T. Yoshida, T. Ozaki, N. Kozai, F. Sakamoto, T. Nankawa, Y. Suzuki, A.J. Francis, Environmental Science & Technology, 41, 3134-3139, 2007. [5] T. Ohnuki, T. Ozaki, T. Yoshida, F. Sakamoto, N. Kozai, E. Wakai, A. J. Francis, and H. Iefuji, Geochim. Cosmochim. Acta, 69, pp. 5307–5316, 2005. [6] T. Ohnuki, T. Ozaki, T. Yoshida, N. Kozai, T. Nankawa, F. Sakamoto, T. Sakai, Y. Suzuki, A. J. Francis, Chem. Geology, 253, pp. 23-29, 2008. [7] T. Ohno Y, Muramatsu, Y, Miura, K. Oda, N, Inagawa, H. Ogawa , A.Yamazaki, C. Toyama, M. Sato, Geochem J 46:287–295(2012). [8] K. Tanaka, Y. Takahashi, A. Sakaguch, M. Umeo, S. Hayakawa, H. Tanida, T. Saito, Y. Kanai, Geochem J 46:73–76(2012) [9]H. Kato, Y. Onda, M. Teramage J Environ Radioact 111, 59–64 (2012) [10] N. Kozai, T. Ohnuki, M. Arisaka, M. Watanabe, F. Sakamoto, S. Yamasaki, M. Jiang, J. Nucl. Sci. Technol., 49, 473-478(2012). [11] T. Ohnuki, N. Kozai, J. Nuclear Science & Technology, 50 (2013) 369-375. 40
  • Development of negative Ce anomalies in biogenic Mn oxide: the role of microorganism on REE mobility during the bio-oxidation of Mn2+ Qianqian Yu a*, Toshihiko Ohnuki a, Kazuya Tanaka b, Naofumi Kozai a, Shinya Yamasaki a, Fuminori Sakamoto a, Yukinori Tani c a Advanced Science Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan b Institute for Sustainable Sciences and Development, Hiroshima University, 1-3-1 Kagamiyama, Higashi-Hiroshima 739-8530, Japan c Institute for Environmental Sciences, University of Shizuoka, 52-1 Yada, Shizuoka 422-8526, Japan * Corresponding author: yu.qianqian@jaea.go.jp. Bio-oxidation of Mn2+ is ubiquitously occurring in most aquatic and terrestrial environments. This dynamic process strongly affected the migration of trace elements as well as radionuclides under environmental condition. On the one hand, Mn oxide, a well-known scavenger of trace elements was formed through enzymatic oxidation. On the other hand, microorganisms released organic matter, which may form complexation with radionuclides [1,2]. Ce is the only rare earth elements (REEs) that can occur in tetravalent state under oxic conditions. Therefore, the fractionation of Ce compared to other REEs provides information about the relative mobility of tetravalent actinides compared to trivalent actinides [3]. In the present work, sorption kinetics of REEs during bio-oxidation and precipitation of Mn2+ was investigated using Acremonium sp. strain KR21-2 under neutral pH conditions in the presence of citrate. Ce anomalies were calculated at nearly zero over 1 hour to 48 hours. A negative Ce anomaly was developed after 72 hours, when biogenic Mn oxide started to form. This result is in contrast to the sorption of REEs on abiotic Mn oxide, which shows clearly positive Ce anomaly. It is well known that Mn oxide oxidize Ce(III) to less soluble Ce(IV), resulting in positive Ce anomaly. Observed Ce anomalies in biogenic Mn oxide are opposite to that in abiotic Mn oxide, indicate that preferential complexation of Ce(IV) with organic matter that released from the organism may change the solubility of Ce(IV) in Mn oxide. We examined the influence of citrate concentration on Mn oxidation rate, REE patterns, and Ce anomaly. The oxidation of Mn2+ was enhanced with small amount of citrate and retarded when citrate concentration is higher than 0.4 mM. The sorption ratio of REE decreased with increasing citrate concentration, but more negative Ce anomaly was observed for the sample without citrate, and in citrate concentration higher than 0.4 mM. In both cases, the oxidation of Mn2+ was suppressed. The trend observed for our data suggested that the mobility of Ce(IV) is highly affected by the relative concentration of organic ligands and mineral surface site. References: [1] K. Tanaka, Y. Tani, Y. Takahashi, M. Tanimizu, Y. Suzuki, N. Kozai, T. Ohnuki, Geochim Cosmochim Ac, 74 (2010) 5463-5477. [2] T. Ohnuki, T. Ozaki, N. Kozai, T. Nankawa, F. Sakamoto, T. Sakai, Y. Suzuki, A.J. Francis, Chem Geol, 253 (2008) 23-29. [3] A. Loges, T. Wagner, M. Barth, M. Bau, S. Göb, G. Markl, Geochim Cosmochim Ac, 86 (2012) 296-317. Fig. 1 The effect of citrate on the sorption ratio of REEs during the formation of biogenic MnO2. 41
  • New Biotechnology Methods for Radioactive Wastes Treatment Alexey Safonov*a, Varvara Tregubova a, Konstantin German a, Olga Gorbunovab, a Frumkin Institute of Physical Chemistry and Electrochemistry Russian academy of sciences, 31, Leninsky prospect, Moscow, Russia 1 b FSUE “RADON”, 2/14, 7th Rostovsky lane, Moscow, Russia, * alexeysafoof@gmail.com Except radionuclides radioactive wastes contain macrocomponents: nitrate ions, acetate, sulfates, TBP and oils. Processing of liquid nitrate and oil-containing radioactive waste to prepare it in a form, suitable for the environmentally safe long-term storage or disposal, is one of the important problems of modern nuclear energetics. A significant number of vacuum and transformer oils are accumulated in nuclear power plants, nuclear fleet and facilities of the nuclear fuel cycle. The problem of oil-containing radioactive waste from radio-chemical industry is quite urgent for several reasons. One of them is complication of processing and extraction cleaning by the presence of the organic phase with a non-permanent composition. Their storage is not safe due to fire risk; cementing is complicated by matrix low oil content (5-7% wt.), and by its possible microbial degradation due to the impact of biogenic organic acids. The main problem of nitrate-containing wastes is their high concentration (1-350 g/l). There are some problems with storing in liquid form in underground repositories because of the presence of highly migrative compounds in solutions. Microbiological destruction is typical for oil- and nitrate-containing cement compounds of long-term storage by the reaction of biogenic organic acids with cement matrix minerals. Biodegradation of LRW components is reasonable before solidification because it reduces volume of LRW and prevents destruction of inorganic cement matrix. The possibility of preliminary microbiological treatment of oil- and nitrate-containing liquid radioactive waste before solidification in the cement matrix was investigated. Biodegradation of oil-containing LRW is possible when using radio resistant microflora for oxidation of oil organic components to carbon dioxide and water. Biosorption of radionuclides by bacteria, followed by oil emulsification in cement solution by biosurfactants (surface-active substances of glycolipid nature) was observed. After 7 days of biodegradation decrease in the volume of oil-containing LRW is found to be reduced because of biodestruction of organic phase to nonradioactive gases (H2O, CO2 and N2), which are removed from LRW volume. Oil destructors bacteria of the genera Pseudomonas, Flavobacterium, Acinetobacter, Aeromonas, Arthrobacter, Rhodococcus, were isolated from low-temperature oil reservoirs and deep storage of liquid radioactive waste. After culturing in the laboratory for a month in a mineral medium with vacuum oil VM-4 with volume ratio 1:10, gravimetric analysis with pre-oil extraction of dichloromethane 5 cm3 showed that the mass of the oil fraction is reduced to 67%. Chromato-mass spectrometry analysis showed a declining of the n-alkane fraction to 50%, izoalkene to 20-25%, and aromatic to 40-60%. Biodegradation of nitrate-containing LRW is possible when using radio resistant microflora for denitrification of nitrate-ions which are the main component of LRW. The process is realized by enzymes catalyzing cellular respiration with consecutive reduction of nitrate-ions to molecular nitrogen through stages of nitrite, nitrous and nitric oxide forms. After 3 days of biodegradation, the concentration of nitrate is found to be reduced from 4 g/dm3 to 10-15 mg/dm3. At the same time by sorption processes at cellular structures microorganisms are able to remove 80-90% of α-radionuclides, up to 50% 90Sr and 20% 137Cs from LRW. Radionuclide containing biomass has to be dried up and solidified into cement matrix. The work is supported by President grant МК-2330.2012.3, RFBR project № 12-08-3127412, and federal program «Kadry» № 2012-1.2.2-12.000-2007 42
  • Removal of Radioactive Cesium from Soil and Sewage Sludge contaminated by Fukushima Daiichi NPP Accident Kenji Takeshitaa*, Yasuhiro Jinbob, Akira Ishidoc, a Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 157-8550 Japan b CDM Consulting Co.LTD, 1-13-13 Tsukiji Chuo-ku Tokyo 104-0045, Japan c Radwaste and Decommissioning Center, 1-7-6 Toranomon, Minato-ku, Tokyo 105-0001, Japan * Corresponding author: takeshita@nr.titech.ac.jp. 1. Introduction Forests, soil and water on the area around the Fukushima Dai-ichi NPP were contaminated widely by the fallout of radioactive Cs. Environmental clean-up techniques to remove and dispose of radioactive Cs from contaminants should be developed immediately for the restoration of Fukushima. However, large amounts of various contaminants make it difficult to attain early environmental restoration. In this paper, a new decontamination technique combined with hydrothermal blasting and the coagulation settling is proposed. The removal of Cs from real contaminants (soil and sewage sludge) sampled in the contaminated area is tested and the applicability of this technique to the environmental restoration is discussed. 2. Combined Process of Hydrothermal Blasting and Coagulation Settling A combined process of hydrothermal blasting and coagulation settling was developed by Tokyo Institute of Technology, CDM consulting Co., and Radwaste and Decommissioning Center. The procedure of the combined process is shown in Fig.1. In the hydrothermal decomposition process, water is added to Cs-contaminants and the water-solid mixture is stirred under the subcritical water conditions (temperature range: 200 to 260ºC, pressure range: 2 to 4 MPa). The ionic product of water increases with increasing temperature and pressure. At 260ºC it reaches about 1000 times that of water at 25ºC. The decomposition of contaminated substances is promoted by the increase in the ionic product of water. After the hydrothermal operation, the high pressure is decreased rapidly by opening a valve. This operation with rapid pressure drop is called “blasting”, by which crystal substances including in the solid can be destroyed partially. The treated water including hydrophilic decomposition products is transferred to the coagulation settling process. The remained solid is washed by water and the washing water is also transferred to the coagulation settling process. Small amounts of coagulant and adsorbent powders are added to the water. Cs ion in the water is recovered as a precipitate. 43
  • Calcination Coagulation settling 44 Washing Hydrothermal 3. Removal of Cs from contaminated Soil and Sewage Sludge The removal of Cs from real sewage sludge (100,000 Bq/kg) sampled in Fukushima-city was tested to demonstrate the validity of the combined process. 30g of the sewage sludge (about 3000 Bq) and 90g of water were added into a 500mL test tube in a hydrothermal decomposition device. The hydrothermal operation was carried out under the conditions of 260ºC and 4.4 MPa for 30 min. The solid remained in the test tube was washed by about 430g of water. Cs in the water was precipitated by the coagulation settling process using ferric ferrocyanide Water 400 g powder (1wt%) as an adsorbent and 260ºC, 4MPa 60 min Solid ion-reaction N (0.1 wt%) produced by NPO Solid Soil 14.2 g 18.5 g 30.0 g 340 Bq Saisei-sha as a coagulant. Fig.2 shows the 690 Bq 1650 Bq Solid Sludge Water results of mass balance and radioactivity 385.3 g 32 g 3~ 6g 1210 Bq 380 Bq Water (Estimate) Water balance. About 72% of the initial 90.1 g 68.6 g Disposal Water radioactivity of sewage sludge was 960 Bq 417 g transferred to water by the simple 40 Bq Water Recycling hydrothermal operation. The final removal Ferric Ferrocyanide 0.2 % Inorganic Coagulant 0.2% of Cs from the sewage sludge was evaluated to be about 96% and most of Cs Fig.3 Decontamination of soil by the combined process with hydrothermal decomposition and coagulation- settling in the sewage sludge was recovered as precipitate. Organic components in the sewage sludge were decomposed to hydrophilic products such as organic acids and dissolved in water. Cs was recovered perfectly from the dirty water containing hydrophilic organic acids by the coagulation settling technique. Next, the Cs removal from contaminated soil (55,000 Bq/kg) sampled at Iidate village, Fukushima prefecture, was tested. 50g of contaminated soil and 150g of water were added into the test tube set in the hydrothermal decomposition device and the soil was decomposed hydrothermally under the conditions of 260ºC and 4.4MPa for 30 min. Fig.3 shows the mass balance and the radioactivity balance. Only 56% of the radioactivity of the soil was transferred to water by the simple hydrothermal decomposition. When the soil remained after the hydrothermal decomposition was washed by a large quantity of water, which corresponded to 13 times the weight of initial soil, the radioactivity transfer of Cs to water was increased to 75%. However, a part of Cs is adsorbed strongly in clay components containing in the soil, such as vermiculite, kaolinite and monmorillonite, and cannot be desorbed by the simple hydrothermal operation. To further improve the removal of Cs from the soil, we tried the blasting operation. After a milder hydrothermal operation (200ºC and 2 MPa for 10 min) the blasting operation was carried out. In spite of milder hydrothermal conditions rather than those in Fig.3, the transfer of Cs to water was increased to 85%. It seems that a part of crystalline structure of inorganic material was destroyed by the rapid pressure drop and a part of Cs adsorbed strongly was transferred to water. The combined process of hydrothermal blasting and coagulation settling is a powerful tool for the removal of Cs from soil.
  • Synthesis of Multifunctional Silica-based Adsorbents and Their Application in Decontamination of Radioactive Contaminated Wastewater Yan Wua, Zi Chena, Qilong Wanga, Hao Wua, Yuezhou Weia,*, Hitoshi Mimurab Shanghai Jiao Tong University, 800 Dongchuan Rd., Shanghai 200240, China b Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2, Sendai, 980-8579, Japan * Corresponding author: yzwei@sjtu.edu.cn a Large amounts of radioactive contaminated wastewater (RCW) were generated from the nuclear accident of Fukushima NPP-1 caused by the Great East Japan Earthquake. The RCW is resulted from the contact of cooling water with nuclear fuel, so the main radionuclides are water-soluble Cs(Cs-134, Cs-137), Sr(Sr-90) and I(I-131, I-129). As seen in the RCW of Fukushima, it contains various co-existent components from sea water (saline elements), corrosion products and ground water. Recently, there are 65 kinds of nuclides such as Pu, MA(Am, Cm) and some FPs( Ru, Rh, Co, etc.) are detected in the RCW. The composition and property of RWCs are significantly complicate so that their decontamination is very difficult. As we know, adsorption method is considered as the key-technology for decontamination of RCW, since it can efficiently remove many kinds of radionuclides even from low concentration of wastewater. The objective of this study is to develop highly functional adsorbents for establishing an efficient and cost-effective treatment system of RCW. In this study, we have prepared different types of porous silica-based inorganic adsorbents and organic/inorganic hybrid adsorbents for decontamination of RCW. These multifunctional silica-based adsorbents have a number of advantages such as fast adsorption kinetics, high uptake capacity, high chemical and radiolytic resistance and ease of solid-liquid separation by simple equipments. The silica-based inorganic adsorbents were prepared by loading a functional compound such as insoluble ferrocyanides, heteropolyacid or titanate into macroporous SiO 2 support by precipitation or crystallization method for selective separation of Cs and Sr (Fig. 1). Silver halides such as AgCl or AgBr were loaded into SiO 2- P (silica/polymer composite support) support for adsorption of I. The organic ligands (TODGA, HDEHP) were impregnated into SiO 2 -P for separation of TRU elements. The adsorption behaviors of Cs, Sr, I and TRU from simulated wastewater were studied and their adsorption kinetics, adsorption capacity and uptake efficiency were evaluated. Fig.1 Schematic preparation procedure This work was financed by the of the porous silica-based inorganic salt National Natural Science Foundation of adsorbents China (21261140335), which is gratefully acknowledged. 45
  • Remove uranium and fluorine from wastewater Gaoyang Ye, Hanyu Wu, Wangsuo Wu* Radiochemistry Laboratory, Lanzhou University, Lanzhou, 730000, China * Corresponding author: wuws@lzu.edu.cn. Uranium conversion factory is an important part of Nuclear Fuel Cycle[1]. The wastewater from the uranium conversion factory contains much uranium and fluorine. Base on correlation laws and request from Project Parties, the effluent can only contain no more than 20 ug/L uranium and 10 mg/L fluorine. In this work, ion exchange and precipitation are used to remove uranium and fluorine from wastewater[2-4]. The uranium and fluorine exist as UO 2 (CO 3 ) 3 4- and F-,respectively. The uranium and fluorine content of the wastewater is about 246 mg/L and 15g/L, respectively. So that we use anion exchange to treat the wastewater hoping remove the majority of uranyl and use precipitate to precipitation the rest of uranyl. Precipitation and sedimentation are effective enough to remove fluorine. We choose 201×7 anion exchange resin because it is cheap and widely used in industry. First 201×7 resin adsorption equilibrium was studied , we found that this resin adsorption uranyl well and the rate is relatively quick. Second the breakthrough curve was investigate. Simultaneously, we also studied the effect and influences on fluorine removal of high fluoride-containing wastewater with chemical coagulation and precipitation process. In this experiment, the concentration of fluorine ion reached 15g/L which is higher than general wastewater. The process combined with CaCl2 and quicklime achieved quite good removal efficiency, and it can be up to the first grade of the National Standard for wastewater discharge (10mg/L), which the general methods cannot reach. The results showed that if we use CaCl2 solely, precipitation cannot subside easily and the residual fluoride ion concentration would be over 200mg/L. However, CaCl 2 combined with quicklime results in better fluoride removal effects. Especially, when the quality of CaCl2 is thrice the quality of quicklime, the fluorine removal is most effective, and the residual fluoride ion concentration is blow 10mg/L which means the removal rate of fluoride ion as high as 99.9%. Compared with general fluoride removal process the advantage of this method is that we can reduce the fluoride ion concentration of high fluoride-containing wastewater to the first grade of the National Standard for wastewater discharge all at once, and multipole chemical technology combination is unnecessary. Table 1. 201×7 resin adsorption equilibrium resin weight/mg ρ(U)/(mg/L) ω R (U)/(mg/g) K d /(L/g) 70.80 63.45 15.47 0.24 90.60 47.10 13.17 0.28 110.30 59.72 10.13 0.17 130.00 54.56 8.84 0.16 150.60 43.46 8.07 0.19 170.30 42.13 7.18 0.17 46
  • Fig.1 201×7 resin breakthrough curve Reference: [1] R.M. Dell, V.J. Wheeler, Transactions of the Faraday Society, 58 (1962) 1590-1607. [2] D.H. Phillips, B. Gu, D.B. Watson, C.S. Parmele, Water Research, 42 (2008) 260-268. [3] F. Semnani, Z. Asadi, M. Samadfam, H. Sepehrian, Annals of Nuclear Energy, 48 (2012) 21-24. [4] F.P. Carvalho, J.M. Oliveira, M. Malta, Ecological Engineering, 37 (2011) 1058-1063. 47
  • Irradiation Stability of the Tributyl Phosphate Solvent Extraction System Cao Xiaominga, Gao yanga,b*, Zheng weifangb a Harbin Engineering University, Heilongjiang Province Harbin, China; b China Institute of Atomic Energy, Beijing, China gaoyang@hrbeu.edu.cn In the Purex spent fuel reprocessing process, tributyl phosphate (TBP) solvent extraction system becomes degraded due to radiolytic and chemical attacks, resulting in a series of temporary degradation products and permanent degradation products. Through alkali washing, the former can be removed but the latter are difficult to be removed from the solvent. These degradation products can cause various physical and chemical damage, and the extraction characteristics of the solvent become deteriorated, leading to impairment of the phase separation, decrease of the mass transfer coefficient of uranium and plutonium, and retention of the fission products in the uranium and plutonium streams. In order to control and optimize the technological conditions of reprocessing process and reduce the irradiation damage of the solvent system, gamma and alpha irradiation stabilities of TBP/n-dodecane system have been studied in this thesis, in which 60Co and 238Pu are adopted as gamma and alpha irradiation sources respectively. The main degradation products dibutyl phosphate (DBP) and monobutyl phosphate (MBP) have been detected by the gas chromatography-mass spectrometry (GC-MS), and the radiation degradation degree of the solvent extraction system has been measured by the test of Pu retention. The effects of the irradiation dose, the concentration of pre-equilibrium nitric acid, the accumulative irradiation dose, and the existence of metal ion on the extraction behaviors of the system have been systematically investigated. The experiment results indicate that: (1) In the range of the absorbed dose studied, the yields of DBP, MBP and the Pu retention of the system all increase with the increase of the absorbed dose after gamma radiation; the pre-equilibration nitric acid concentration has a significant effect the yield of DBP and the Pu retention, but has little impact on the output of MBP; the more the number of irradiation, the larger the Pu retention value. Moreover, the Pu retention of the system by accumulated radiation is greater than that by a single radiation under the similar absorbed dose. (2) In the range of the absorbed dose studied, the Pu retention of the system increased with the increase of the absorbed dose after alpha radiation. When the absorbed dose is 1×106Gy, the Pu retention approaches 40%; the effect of the pre-equilibration nitric acid concentration on the Pu retention is not obvious; the existence of Zr4+ in the solvent system intensifies irradiation damage of the system, and the Pu retention decreases monotonically with the increase of the concentration of Zr4+. [1] I.A. Kulikov, N.V. Kermanova and M.V. Vladimirova, Sov. Radiochem. 25, 310 (1983). [2] I. A. Kulikov, N. V. Kermanova, O. A. Sosnovskii, N. N. Shesterikov and M. V. Vladimirova, Radiokhimiya 23, 825 (1981). [3] B. G. Brodda, and D. Heinen, Nucl. Technol., 34, 428-437 (1977). [4] Z. Nowak, M. Nowak and A. Seydel, Radiochem. Radioanal. Lett. 38, 343 (1979). [5] T. Ladrielle, P. Wanet, D. Lemaire and D. J. Apers, Radiochem. Radioanal. Lett. 59, 355 (1983). 48
  • 5x105 5x105 4x105 4x105 5x105 DBP DBP MBP MBP DBP MBP 5 4x10 峰面积 peak area 峰面积 3x105 3x105 2x105 2x105 1x105 1x105 3x105 2x105 1x105 0 0 0 -1x105 -1x105 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106 0.0 0.02.0x105 4.0x105 6.0x105 8.0x105 1.0x106 吸收剂量(Gy) absorbed dose(Gy) 1 Fig. 1. Effect of the gamma absorbed dose on the amount of DBP and MBP in 30%TBP-n-dodecane 2 3 4 预平衡硝酸浓度(M) 5 Fig. 2. Effect of the concentration of nitric acid on the amount of DBP and MBP in 30%TBP-n-dodecane 0.5 0.5 0.4 0.3 Pu retention Pu retention 0.4 0.2 0.3 0.2 0.1 0.1 0.0 0.0 0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106 0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106 cumulated absorbed dose (Gy) absorbed dose Fig. 3. Effect of the gamma cumulated absorbed dose on the Pu retention in 30%TBP-n-dodecane Fig. 4. Effect of the alpha absorbed dose on the Pu retention in 30%TBP-n-dodecane 0.4 Pu retention 0.5 0.4 Pu retention 0.5 0.3 0.2 0.3 0.2 0.1 0.1 0.0 0.0 1 2 3 4 concentration of nitric acid (M) 0.3 5 0.6 0.9 1.2 1.5 concentration of Zr4+(mg/ml) 1.8 Fig. 6. Effect of the existence of UO22+ ion on the Pu retention in 30%TBP-n-dodecane Fig. 5. Effect of the concentration of nitric acid on the Pu retention in 30%TBP-n-dodecane 49
  • U(VI) sorption on silica in the presence of short chain mono-carboxylic acids a Yujia Zhao a *, Florence Mercier-Bionb, Grégory Lefèvrec, Eric Simonia UMR 8608, Institut de Physique Nucléaire d’Orsay, Université Paris-Sud 11, 15 rue Georges Clémenceau 91406 Orsay Cedex, France b Laboratoire Archéomatériaux et Prévision de l’Altération, Centre CEA Saclay 91191 GIF-SUR-YVETTE Cedex, France c UMR 7575, Chimie ParisTech, Ecole Nationale Supérieure de Chimie de Paris, 11 rue Pierre et Marie Curie 75005 Paris, France * Corresponding author: yujia.zhao@sjtu.edu.cn This subject comes within the scope of treatment and disposal of nuclear waste. Understanding the migration behaviour of radionuclides is essential for a reliable long-term safety assessment of nuclear waste disposal sites. In this study, we focus on the sorption behaviour of uranyl ion (one of the waste products and model of hexavalent actinides) on silica gel (reference surface of oxides present in soils) in the presence of the simplest monocarboxylic acids (to model the organic matters or to be degradation products of cellulose issued from nuclear industry). Another reason for which we are interested in this system is that no investigation has been reported on interactions of uranyl/formic (acetic, propionic) acid /silica in previous studies, while the main part of studies on ternary systems concerns the effect of humic or fulvic substances. In this work, the studies of uranyl ion and acids uptake in sorption systems have been performed by combining the macroscopic sorption data and the spectroscopic information of the surface complexes. The sorption edges as a function of pH for different systems indicate that the increase of organics concentration results in a decrease of uranyl ion retention in the following order: propionate > acetate > formate, which can be interpreted as their complexing capacity with uranyl ion in solution. Obviously, the presence of carboxylic acids as well as their carbon chain length influence the uranyl sorption profile. To provide the structural information, ATR – FTIR and TRLFS are applied to carry out the speciation of uranyl ion and carboxylic acids at the silica/electrolyte interface. Two techniques show a good agreement that the presence of acids changes the environments of sorbed uranyl. Both techniques suggest the existence of “silica-uranyl-organic” ternary surface complex when acetic or propionic acid presents. The characterization of sorbed species (number, nature, their distribution as a function of pH) is determined experimentally. Based on these structural investigations, the sorption edges are simulated effectively and the reaction constants are then obtained by using the constant capacitance surface complexation model. 50
  • Effect of some ions on the sorption of Th(IV) to K-feldspar Yaofang Du, Ping Li, Zhuoxin Yin, Wangsuo Wu* Radiochenmistry Laboratory, Lanzhou University, Lanzhou 730000, China * Corresponding author: wuws@lzu.edu.cn With the wide application of nuclear energy, thorium resources as a kind of potential nuclear fuel has been widely development and utilization, but the resulting radioactive pollution nots allow to ignore, especially of the waste water containing thorium coming from smelting plants, this radioactive thorium may cycle into the biosphere with water, cause serious harm to the organism’s safety and the ecological environment, so the purification of thorium pollutants in water environment is of great importance. The sorption of Th(IV) on K-feldspar was studied by batch technique in detail as a function of contact time, pH, background electrolyte, HA, FA, and temperature, The results indicated that a rapid sorption equilibrium was achieved in less than 8 hours, and the dynamic of sorption was simulated well by the Pseudo-Second-Order model; The sorption was strongly dependent on pH; monovalent cations (K+, Na+, Li+) and anions (ClO 4 -, NO 3 -, Cl-) had negligible effects on Th(IV) sorption; PO 4 3dramatically enhanced sorption by forming ternary complex on K-feldspar surface, while SO 4 2inhibited sorption; With growing concentration of HA/FA, sorption of Th(IV) was promoted markedly, and the positive effect of FA was much stronger than that of HA. High temperature was favorable for Th(IV) sorption on K-feldspar, indicating sorption process was endothermic. Sorption was fitted better by Langmuir model than Freundlich model. 80 60 60 sorption% 100 80 Sorption% 100 40 NaClO4 NaCl NaNO3 20 0 2 3 4 5 6 7 8 40 NaH2PO4 NaClO4 20 NaHSO4 0 9 1 pH 2 3 4 5 6 7 8 9 pH Fig.1 Effect of some negative ions(I) on Fig .1 Effect of SO 4 2- and PO 4 3- on the the sorption of Th(IV) to K-feldspar sorption of Th(IV) to K-feldspar C[Th4+]in=8.33×10-5 mol/L, T=298.15±1 K, m/V=0.6 g/L, I=0.01 mol/L C[Th4+]in= 8.33×10-5 mol/L, T=298.15±1 K, m/V=0.6 g/L, I=0.01 mol/L 51 10
  • URANYL IONS SORPTION TO TIO 2 AND INTERACTION WITH SORBED FA: EXPERIMENTS AND MODELING Yuanlv Ye, Xuefeng Wang, Ning Guo, Zhijun Guo, Wangsuo Wu* Radiochemistry Laboratory, Lanzhou University, Lanzhou, 730000, China * Corresponding author: wuws@lzu.edu.cn In the environment, an important factor of radionuclide mobility is their interaction with mineral– water interfaces. To predict radionuclide mobility, it is necessary to understand fundamental processes such as surface precipitation and surface complexation. Studies of uranium sorption onto mineral surfaces have great practical importance for risk assessment1-3 .In recent years it turned out that many groundwater in granitoidic environment contain excessive amounts of dissolved U 4. In this work, experiments and modeling studies are performed to elucidate the interaction of fulvic acid (FA) with uranyl ions in the presence of TiO 2 surfaces. FA is strongly bound to TiO 2 , and has a very strong effect on the U(VI) sorption. U(VI) sorption to TiO 2 in the presence and absence of sorbed FA can be well predicted with the SCD model (surface and complex distribution). According to the model calculations, the nature of the interaction between FA and U(VI) at the TiO 2 surface is mainly surface complex formation. This is the first time that effects of natural organic matter (NOM) on the sorption of a cation are predicted successfully using an integrated ion-binding model for oxides and for humics that accounts for the chemical heterogeneity of NOM. 100 FA sorption % 80 60 40 no U(VI) C(U(VI))=4.012E-4mol/L C(U(VI))=8.024E-5mol/L 20 0 2 3 4 5 6 7 8 9 pH Fig. 1 Sorption of FA onto TiO 2 vs. pH in the absence and presence of FA, C [FA] =20 mg/L, m/ V = 20 g/ L , T = 22±1 ℃, I =0.1 mol/ L (NaCl). References: [1] Van Loon, L. R.; Baeyens, B.; Bradbury, M. H. Appl. Geochem. 2009, 24, 999. [2] Rabung, T.; Pierret, M. C.; Bauer, A.; Geckeis, H.; Bradbury, M. H.; Baeyens, B. Geochim. Cosmochim. Acta, 2005, 69, 5393. [3] Bradbury, M. H.; Baeyens, B. Geochim. Cosmochim. Acta, 2005, 69, 5391. [4] Gustafsson, J. P.; Dassman, E.; Backstrom, M. Appl. Geochem. 2009, 24, 454. 52
  • Thermal Decomposition Behavior of Nitrate Solution containing Di-n-butylephosphate in Vitrification Process Tatsuya Fukudaa, Yoshio Nakanoa, Kenji Takeshitaa* Kazuhiro Minamib, Eiji Ochib a Research laboratory for Nuclear Reactor, Tokyo Institute of Technology b Japan Nuclear Fuel Ltd. takeshita@nr.titech.ac.jp 1. Introduction The cold cap formation floating on top of the molten glass pool plays an important role in order to stably operate vitrification process. The formation of the cold cap and its conversion to molten glass take place under non-isothermal conditions and depend on properties of the various chemical elements and many processing parameters. Especially, the presence of di-n-butylephosphate that comes from degradation of TBP extractant has been known to govern the cold cap formation significantly. However, the mechanism of cold cap formation in the presence of DBP has not been clarified because of its complexity. In this study, the thermal decomposition behavior of nitrate solution containing DBP was studied in detail. 2. Experimental Zr nitrate, which has high affinity for DBP in HLLW, was chosen among major nitrate. Zr-DBP complex was prepared by slowly adding 1.0w% of DBP to 0.1 M ZrO(NO3)2・2H2O in 3 M HNO3 solution. Aggregated Zr-DBP complex was filtered and dried at room temperature. Thermal decomposition of ZrO(NO3)2・2H2O and Zr-DBP complex was studied by TG and TG-DTA-MS. The composition of Zr-DBP complex decomposed at 1000 ℃ was determined by Xray diffraction analysis. 3. Results・Conclusion Dried Zr-DBP complex was white in color and sticky. Thermal decomposition behavior of the aggregation and ZrO(NO3)2 ・2H2O was shown in Fig. 1. The reaction rate (dα/dt [1/sec], α: Conversion ratio) vs temperature was shown in Fig. 2. These figures show that thermal decomposition behavior of ZrO(NO3)2 ・2H2O and Zr-DBP complex are completely different from each other. Reaction path in temperature rising condition is found as follows, Zr nitrate: ZrO(NO3)2→Zr2O3NO3→ZrO2 Zr-DBP complex: Zr-DBP complex→…→ZrP2O7 It was clarified that phosphoric group remains in the heated product of Zr-DBP complex, even in 1000 ℃. In conclusion, DBP in HLLW was found to play an important role to form cold cap. Fig. 1. TG curve of Zr compounds Fig. 2. Reaction rate of Zr compounds (This work is a part of the research supported by Japan Nuclear Fuel Limited with Grant-in-Aid by the Ministry of Economy, Trade and Industry.) 53
  • Study on the synthesis of AMP loaded silica and its adsorption behavior for Cs Qilong Wanga, YanWua*, Yuezhou Weia, Hitoshi Mimurab a Shanghai Jiao Tong University, 800 Dongchuan Rd., Shanghai 200240, China b Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2, Sendai, 980-8579, Japan Corresponding author: wu_yan@sjtu.edu.cn After the FUKUSHIMA nuclear accidents, much more attentions have been given to the removal of 137Cs from the radioactive contaminated wastewater. 137Cs with a half-life of 30 years exhibits high radioactivity and heat generation. Therefore it’s of great importance to remove Cs from the radioactive contaminated wastewater for the environmental safety. The Ammonium molybdophosphate (AMP), which shows high selectivity towards Cs+ ion, is reported to be a promising adsorbent to remove Cs in radioactive contaminated wastewater [1, 2]. AMP exhibits high ion exchange capacity,excellent heat and radiation resistance. However the AMP is not used in large scale for its fine powder form, which hinders the simple column operation. To solve this problem, in this study, the AMP particle was loaded onto a porous silica support by crystallization method to synthesize a new adsorbent AMP/SiO2. The new synthesized adsorbent has many advantages, such as fast adsorption kinetics, excellent mechanical and chemical stability,and more beneficial for the column operation. The surface morphology of AMP-SiO2 was examined by scanning electron microscopy (SEM) , the adsorbed metal ions on the AMP-SiO2 was examined by energy dispersive spectroscopic (EDS). Batch experiment was carried out to test the uptake speed for Cs of AMP-SiO2 with different kinds of SiO2 support including SiO2-F50(particle size:50µm, pore size:50nm) ,SiO2-F100(particle size:50µm, pore size 100nm) and SiO2-F200(particle size:50µm,pore size:200nm).To make a comparison the same experiments were also conducted using zeolite adsorbents.Column operation was conducted using AMP/SiO2-F50 to get the breakthrough curve and calculated the adsorption capacity. In this study, a spherical AMP-SiO2 was obtained by SEM. It takes just 10 min for the AMP-SiO2 adsorbents to attain equilibrium, and the uptake percentages nearly 99% of Cs+ were obtained in the presence of 0.6 M NaCl (Fig. 1). The order of initial uptake percentage is F50>F100>F200. On the other hand, it takes more than 48h for the zeolite adsorbents to uptake no more than 65% of Cs+(Fig. 2).For column experiment, an S-shaped breakthrough curve was obtained with a total adsorption capacity calculated to be 0.33mmol/g. The AMP-SiO2 absorbing Cs+ was analyzed by EDS(Fig. 3), from the cross section of AMP-SiO2, we can detect the spectrum of Cs together with that of Mo, this indicated that AMP-SiO2 has high selectivity toward Cs. This work was financed by the National Natural Science Foundation of China (21261140335), which is gratefully acknowledged. [1]. Hitoshi Mimura, Yan Wu, Wang Yufei, Yuichi Niibori, Isao Yamagishi, Masaki Ozawa, Takashi Ohnishi, Shinichi Koyama, Selective separation and recovery of cesium by ammonium tungstophosphate-alginate microcapsules, J. Nucl. Eng. Des., 241, 12: 4750-4757( 2011). [2].Yan Wu, Hitoshi Mimura, Yuichi Niibori, Takashi Ohnishi, Shinichi Koyama, YueZhou Wei, Study on Adsorption Behavior of Cesium Using Ammonium Tungstophosphate (AWP)-Calcium Alginate Microcapsules, Science China Chemistry, 55(9):1719-1725( 2012). 54
  • Fig. 1 Effect of shaking time on uptake(%) of Cs for AMP-SiO2.V/m=400cm3/g, [Cs]:20ppm, [NaCl]:0.6M, 25oC. Fig.2 Effect of shaking time on uptake(%) of Cs for zeolite.V/m=400cm3/g, [Cs]:20ppm, [NaCl]:0.6M, 25oC. Fig.3 EDS analysis of AMP-SiO2 after adsorption of Cs. 55
  • Abstract for ASNFC 2013 Selective Adsorption and Stable Solidification of Sr by Potassium Titanates Hitoshi Kandaa, Hitoshi Mimuraa, Yuichi Niiboria, Mamoru Iwasakib, Kouichi Morib, Nobuki Itoic, Toshiki Gotoc a Tohoku University, b Kurita, c Otuka Chemical * Kanda: candy@michiru.qse.tohoku.ac.jp The development of selective adsorbents for radioactive Sr is one of the most important subjects for the safety decontamination in Fukushima NPP-1[1]-[4]. In this paper, selective adsorption properties of Sr, characterization and stable solidification were studied by using the novel adsorbent of potassium titanates (KT-1). The adsorption properties of Sr for original and the calcined specimens were examined by batch method under the following conditions; V/m= 100 cm3/g, Mixed solution: 10,000 ppm Na+, 10 ppm Cs+, 10 ppm Ca2+, 1 ppm Mg2+ and 1 ppm Sr2+, 85Sr tracer: 5,000 cpm/cm3, centrifugation: 2,500 rpm, 25℃, shaking time: 1~24 h, calcination temp.: 300~900℃. Relatively large uptake percentage above 90% was obtained for the original and calcined specimens below 800℃, while the Sr uptake for the calcined specimens above 900℃ was lowered due to the thermal decomposition of K 2 Ti 2 O 5 ・xH 2 O (Fig. 1). The Sr distribution in the column was examined by flowing the mixed solution through the columns packed with KT-1. The Sr distribution profiles were obtained by the measurement of γ -activity in the column of 5 mm intervals. In either case, no breakthrough of Sr was observed. The distribution profile tends to smooth with increasing the flow rate; Sr adsorption band and flow rate have a linear relationship as shown in Fig. 2. The leachability of Sr for the solid forms was further examined under the following leaching conditions; leachant: pure water and 1 M HCl; leachant temp.: 25℃ and 90℃, leaching period: 4 weeks; calcining temp.: 500∼1,100℃. The leached percentage of Sr in pure water was less than the detection limit of ICP-AES, and that in 1 M HCl tended to markedly decrease with calcining temperature; the formation of SrTiO 3 phase above 800℃ was effective for the lowering of leachability. The novel adsorbent of KT-1 is thus effective for the selective decontamination and stable solidification of Sr in Fukushima NPP-1. Fig. 1 Uptake(%) of Sr for KT-1 calcined different temperatures. Fig. 2 Relation between adsorption band length and flow rate. [3] H. MIMURA, I. YAMAGISHI, “Characterization and adsorption properties of selective adsorbents for high decontamination of cesium” Journal of Ion Exchange, Vol.23, No.1, 6-20 (2012). 56
  • Adsorption and Stable Solidification of Cesium by Insoluble Ferrocyanide Loaded Porous Silica Gels Xiang-biao Yina, Yan Wub, Hitoshi Mimuraa, Yue-Zhou Weib, Yuichi Niiboria Department of Quantum Science and Energy Engineering, Tohoku University, Aramaki-Aza-Aoba 6-6-01-2, Sendai, 980-8579, JAPAN b School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai, CHINA * Corresponding author: yin@michiru.qse.tohoku.ac.jp a In Fukushima NPP-1 accidents, high-activity-level water (HALW) accumulated in the reactor, turbine building and the trench in the facility is treated by circulating injection cooling system. Though the system has been effectively operated and the cold shutdown had been completed, large amounts of solid waste such as zeolites and insoluble ferrocyanide sludge are generated. Hence the development of effective treatment and disposal methods is very urgent and important subject. For the selective adsorption and stable immobilization considering the safety treatment and disposal of secondary solid waste containing radioactive cesium, novel porous silica gels loaded with insoluble ferrocyanide (SLFC) were prepared by impregnation/precipitation method. Base on the previous research results, the SLFC composites have relatively large distribution coefficients of Cs+ ions and adsorption rate constants. The solidification results indicate that zeolites have excellent Cs immobilization characteristic, gas trapping and self-sintering abilities and low leachability. This subject further choose three different kinds of SLFC composites (Q10-NH, Q10-PEI, NiFC-SG) by mixing with nine kinds of additives at higher temperature above 900 ℃ to achieve the optimization of solidification method. The Cs contents in three composites were estimated to be below 4.5 wt% at higher temperature above 1,000℃ and decreased with three stages after calcination (Fig. 1). In contrast, the lowering of Cs immobilization ratio was markedly depressed by mixing additives in which allophane (All) had the best result. By increasing the additive ratio to 1:1, the Cs immobilization ratio was almost 100% and no volatilization of Cs was detected even after calcination at 1,200℃ (Fig. 2),indicating that the calcination of the mixture of (SLFC composites)/(specific additives) at appropriate ratio was thus effective for the stable solidification. [1] H. MIMURA, T. KANNO, J Nucl. Sci. and Technol., 22(4), P. 284-291 (1985) [2] H. MIMURA, K. YOKOTA, K. AKIBA, J Nucl. Sci. and Technol., 38(9), P. 766-772 (2001) 8 Q10-NH Q10-PEI NiFC-SG 6 1:1 Immobilization ratio of Cs ( %) Cs content (wt%) 7 5 4 3 2 1 0 0 200 4:1 9:1 1:0 1200℃ 80 60 40 20 0 400 600 800 1000 1200 Temperature ( ℃) 2:1 100 A X Y Cha Mor Cli Clay Fly All Addictives Fig. 1 Comparison of Cs content (wt%) for different adsorbents Fig. 2 Immobilization efficiency in different mixing after calcination ratios (Mixing ratio = Q10-PEI : additive) 57
  • Abstract for ASNFC 2013 Separation of Nuclides by Different Types of Zeolites in the Presence of Boric Acid Natsuki Fujitaa, Hitoshi Mimuraa, Takaaki Kobayashib, Kazuyuki Sekinob, Kunitaka Nagaminec a Tohoku University, b Mitsubishi Heavy Industries, LTD, c Nuclear development Corporation * Fujita: natsuki@michiru.qse.tohoku.ac.jp The development of selective adsorbents for multi-nuclides has become important for the effective decontamination. In this paper, the selective adsorption properties of 26 nuclides for different types of zeolites (A, L, mordenite, Ag-mordenite) was examined in the presence of boric acids considering the PWR operation. The batch adsorption experiments were carried out using four kinds of test solutions containing boric acid and calcium hydroxide; ①DW(distilled water) + H 3 BO 4 : 3,000 ppm + LiOH: 10 ppb, ② DW + Ca(OH) 2 : 500 ppm + H 3 BO 4 : 3,000 ppm + LiOH: 10 ppb, ③seawater(30% diluted) + H 3 BO 4 : 3,000 ppm, ④Seawater + H 3 BO 4 : 3,000ppm. The uptake (%) of Sr2+ for zeolite A (A-51J), Cs+ for natural mordenite (NM, 2460#, Ayashi, Sendai), and I- for Ag-NM was determined under the following conditions; concentration of Sr2+, Cs+ and I-: 10 ppm, V/m= 100 cm3/g, 25℃, 24 h. The uptake percentage of Sr2+, Cs+ and I- ions was determined to be above 90%, while tended to decrease in the presence of seawater. Especially, the uptake percentage of I- ions for Ag-NM markedly decreased in the presence of seawater (Fig. 1). As for the zeolites A and L, the uptake (%) of 26 elements was determined by using four kinds of test solutions. Zeolite A has relatively large uptake percentage for Sr, Co, Ni and Zn, and zeolite L has high adsorbability to lanthanoid group of Eu, Ce and Pr. The increase in pH led to the enhancement of uptake (%), while the hydrolysis of metal ions should be also considered (Fig. 2). The multi-nuclides separation is thus expected by considering Fig. 1 Uptake (%) of Sr, Cs and I for the difference in uptake properties of zeolite A, zeolite A, NM and Ag-NM. L and natural mordenite. Fig. 2 Uptake (%) of 26 elements for zeolite A (left, pH 4) and zeolite L (right, pH 4). 58
  • Session 4: Transmutation, Resources and Materials Utilization, etc. 59
  • Hydriding Properties of Uranium Alloys—Their Meaning for Nuclear Fuel Cycle Michio Yamawaki a,*,Yuji Arita a, Takuya Yamamoto a ,Fumihiro Nakamori a ,Kazuto Ohsawab a University of Fukui b Kyushu University * Corresponding author: yamawaki@u-fukui.ac.jp Uranium alloys may be useful as hydrogen storage material if powdering during hydriding-dehydriding cycles can be prohibited. UNiAl intermetallic compound may be a candidate for such anti-powdering uranium alloys[1]. Hydrogen absorption properties of UNiAl has been studied experimentally as well as calculationally. Experimentally it was found to absorb and desorb hydrogen without powdering. While, calculationally the change of crystal structure from low hydrogen content A-type structure to high content B-type structure has been evaluated to occur at around 2.5 of H/UNiAl ratio in accord with the experimental result. Also, the calculation using VASP code has revealed the lattice constant of a axis expands with increase of hydrogen content in UNiAlHx, while that of c axis shrinks. Such a trend is in accord with the experimental observation. Long-time endurance test will be needed and acceleration of absorption and desorption rates will be necessary for future practical use of such a compound as hydrogen storage material. The meaning of development of such a uranium-base hydrogen storage material is to alleviate the pressure due to idling storage of large amount of depleted uranium resulting from the enrichment process in the upstream of the fuel cycle. Feasibility of the practical use of this kind of hydrogen storage material is to be discussed. [1] T. Yamamoto, H. Kayano and M. Yamawaki, J. Alloys Compds., 213/214, 533 (1994). Fig. 1. B-type Structure of UNiAlHx Red spot: U, Blue spot: Ni, White spot: Al, Purple square; H (D) 60
  • Microstructural Study of As-Cast U-Rich U- Zr Alloys Y.T. Zhanga,*, P.C. Zhanga, X.Y. Wang a, J.J. Pingb, G. Zengb, X.X. Pangb and Q.Y. Xua a Science and Technology on Surface Physics and Chemistry Laboratory, Mianyang 621907, China b Material Institute of CAEP, Mianyang 621900, China * Corresponding author: zhangyuting03@163.com. Metallic fuel is one of the candidate fuel forms to be used for generation Ⅳ reactors. It has many advantages such as thermal conductivity, evolution under burn-up and other factors[1]. Metallic fuel such as U-Zr alloys have been considered as a nuclear fuel for fast reactor, also U-Zr is the major sub-system of U-Pu-Zr or U-Zr-Nb system for fast breeder reactor fuel. The constituent phases of U-Zr alloy is complicated, it is of great importance to study its phases for investigation of alloy irradiation performance, for metallic fuel having a fine micro-structure results in a higher fission gas release rate during irradiation[2,3]. According to the U-Zr equilibrium phase diagram[4], uranium-rich U-Zr alloys consist of a two-phase structure of orthorhombic α-U and the hexagonal intermetallic δ phase. Conclusive experimental evidence wasn’t provided[5]. The present investigation deals mostly with the microstructure of as-cast seven U-rich U-Zr alloys. XRD and neutron diffraction data indicates that the alloy consists predominantly of the α phase, difficult to discern are peaks that correspond exclusively to the δ phase(U 2 Zr). This study confirms exclusively δ phase existing in as-cast U-rich U-Zr alloys and revealing the orientation relationships. Fig.1 reveals lamellar structure in as-cast various U- Zr alloys, which consists of two phases. Inter-lamellar spacing reduces as Zr content increases in the alloys. Fig.2 shows typical transmission electron micrographs of the U-Zr alloy, showing the morphology of the lamellar structure. The average space of lamellae is ~70nm, the average width of the δ phase (the gray in Fig.2a) is ~50nm. In STEM mode, the brighter in Fig.2b is δ phase, for its thicker. SAD pattern shows in Fig.3a indicating the orientation relationships between the two _ _ _ lamellar phases: (010)[001] α ‖(0110)[21 10] δ . The results is consistent with early study. Fig.3b shows the two phases have coherent interface, there are misfit dislocations on the interface. As-cast U-Zr alloy consist of two phases, α-U and intermetallic δ phase, has the morphology of the lamellar structure, which was_ confirmed exclusive by TEM. They have _ _ orientation relationships: (010)[001] α ‖(0110)[2110] δ . [1] G.L. Hofman, L.C. Walters, T.H. Bauer, Prog. Nucl. Energy. 31 (1997). [2] C.K. Kim, J.M. Park, H.J. Ryu, Nul. Eng. Technol. 371(2007). [3] K.H. Kim, H.J. Kwon, J.M. Park, Y.S. Lee, C.K. Kim, J. Korean Nucl. Soc. 33(2001). [4] ASM Alloy Phase Diagrams Center, P. Villars, editor-in-chief; H. Okamoto, K. Cenzual, section editors; ASM International, Materials Park, OH, 2006. [5] J.T. McKeown, S. Irukuvarghula, S. Ahn, M.A. Wall, L.L. Hsiung, S. McDeavitt, P.E.A. Turchi, Journal of Nuclear Materials, 436(2013). 61
  • (a) (b) (c) (d) (e) (f) Fig.1. Scanning electron microscopy images of as-cast U-rich U-Zr alloys.(a)U-2Zr; (b)U-4Zr (c)U-6Zr; (d)U-8Zr; (e)U-10Zr; (f)U-12Zr. Fig.2. (a) TEM and (b) STEM pictures showing laminar structure. Fig.3. (a) SAD patterns showing the orientation relationship and (b) HREM image showing the coherent interface between the α and δ phases. Arrows point to interface in (b) to highlight misfit dislocations. 62
  • Production of standard particles and their application in particle analysis for nuclear safeguards Li Jin-yinga, Wang Fanb, Chen Yanb, Zhang Yanb, Wang Tong-xingb, Shen Yanb, Zhao Yong-gangb, Chang Zhi-yuanb, Shi Leia a. China Resources Co., Ltd., Beijing 100005, China; b. China Institute of Atomic Energy, P.O.Box 275-8, Beijing 102413, China Novel aspects: Two particle analysis techniques were compared by using standard particles. Introduction: Isotopic ratio in uranium-bearing particles from swipe samples provides important information for detecting undeclared activities in nuclear safeguards. Particle analysis includes particle recovery, identification, location, transfer and isotope measurement. Fission track analysis combined with thermal ionization mass spectrometry (FT-TIMS) and secondary ion mass spectrometry (SIMS) are widespread and adopted by IAEA and its network of analytical laboratories. Methods: The standard particles were produced via aerosol spray pyrogenation and characterized by SEM with EDX. For SIMS, the standard particles were recovered by ultrasonic bath and then searched and measured by SIMS. For FT-TIMS, The particles also were recovered by ultrasonic bath and dropped onto a polycarbonate membrane and Then covered with a FT detector. After irradiation and etching, the uranium oxide particles were able to find and locate under microscope. The uranium oxide particle was picked up by a glass needle and transferred onto a rhenium filament for TIMS measurement. Results: Standard particles, monodisperse uranium oxide microspheres consisting of certified isotopes of 1 μm in diameter, had been produced. The characteristics were confirmed by SEM-EDX. Then these standard particles were used for evaluation of two traditional particle analyses, SIMS and FT-TIMS. The result showed that these techniques were adequate for uranium isotope measurement, SIMS was more efficient and FT-TIMS was more precise. 63
  • Après ORIENT, A New P&T Challenge to Tran smute Radioactive Wastes into Resources Masaki Ozawa Tokyo Institute of Technology, International Nuclear Research Cooperation Center 2-12-1 N1-21, Research Laboratory for Nuclear Reactors, Ookayama, Meguro-ku, Tokyo 152-8550 Japan After the catastrophe of Fukushima Daiichi NPP on March 11, 2011, the sentiment of Japan against nuclear energy has been drastically shifted to be negative mind. However, the fuel cycle roles to maintain sustainable, carbon-emission free energy resource (i.e., 238U-239Pu) and to contribute to energy security are still, or even more become important. In this context, radioactive waste management will require dramatic conceptual change with technological innovation. Après ORIENT research program was newly initiated in 2011 subsequently to Adv.-ORIENT Cycle (since 2006) as shown Fig.1. The Après ORIENT will deal with transmutation of radioactive FP elements to create rare metals / rare earth (RE) elements (i.e., Nuclear Rare Metal, NRM) by (n,γ) reaction with subsequent β- decays, A A+1 → Z+1 NRMA+1 + β Z FP (n,γ ) Z FP The reaction might be called as modern alchemical reaction. The current target Z FPA must be radioactive FPs, and eventually created Z+1 NRMA+1 should be more valuable and strategically important, with non or extremely less radioactive than that of mother Z FPA. In addition, created A+1 must be the Z+1 NRM elements which are Advanced ORIENT Cycle 235U (n,νn) FP , Energy (7.8 x107KJ/235Ug) categorized ones with higher supply risk in the leading‐edge industries. Geological Disposal Non or Very Low Radioactive ⇢Use In the case of The Primary Nuclear Rare Metals ; Conventional Fuel Cycle transmutation of Mo,In,La,Nd,Dy,PGM(Ru?,Rh?,Pd) radioactive Ba (i.e., non High Radioactive RE element) to La (i.e., Create New Rare Metals by light RE), the yield ■ Après ORIENT 139 Neutronic Transmutation of of La was low < 3.5 % Radioactive Wastes with 5 years irradiation at A A+1 → A+1 + βordinary thermal or fast ZFP (n,γ ) Z FP Z+1RM neutron energy condition. Use ⇠ Non Radioactive, High Value and Strategic But, the transmutation The Secondary Nuclear Rare Metals ; rates would be improved PGM(Ru,Rh,Pd,Os?), In,La,Gd,Tb,Dy, Re?, etc to be more than 12 % / 5y provided at the blanket Fig.1 Concepts of Advanced and Après ORIENT Cycle part of fast reactor corresponding to the resonance energy region. Created La will be perfectively non radioactive (<0.001Bq/g) despite that natural La is slightly radioactive ca. 1Bq/g. Adsorption behaviors of the elements constituting high level liquid wastes (HLLW) were 64
  • investigated for newly synthesized crown ether (CE) resins. Among them the highest distribution coefficient (Kd) of 2+ divalent Ba and Sr2+ were observed for Benzo-18-Crown-6 (BC18) - substituted resin in the simulated HLLW / 9M HCl media as shown in Fig. 2. Since little affinity and size dependency were observed for mono-valent cations and tri-valent RE elements, quantitative separation of Fig.2 Adsorption of Elements for Benzo-18-Crown-6 (BC18) Ba2+ from them are substituted Resin; simulated HLLW, 9M HCl medium, ambient highly expected. Another temperature mechanism might involve in the adsorption of oxo-anion, Mo, Re and Tc. To confirm such P&T, proof-of-concept experiments were carried out firstly for light RE group elements at the TRIGA reactor, UC-Irvine, in the collaboration with Tokyo Tech. [1] Masaki Ozawa, Tatsuya Suzuki, Shinichi Koyama, Hiroshi Akatsuka, Hitoshi Mimura and Yasuhiko Fujii, A New Back-end Cycle Strategy for Enhancing Separation, Transmutation and Utilization of Materials (Adv.-ORIENT Cycle), Progress in Nuclear Energy, 50(2008)476-482, ELSEVIER (2008). [2] HAN Chi Young, OZAWA Masaki, SAITO Masaki, Resourceability on Nuclear Fuel Cycle by Transmutation Approach, SCIENCE CHINA Chemistry (SPECIAL TOPIC: Nuclear Fuel Cycle Chemistry), Vol.55 No.9, pp.1746-1751, September 2012. [3] Masaki Ozawa, From Radioactive Wastes to Resources, by Neutronic Transmutation to Create New Rare Metals, Rare Earth, PIM2013 Processes in Isotopes and Molecules, Plenary Pl-11, 25-27 September 2013, Cluj-Napoca, Romania. [4] Masaki Ozawa, Chi Young Han, Toshitaka Kaneshiki, Masao Nomura, Shinichi Koyama, Mikael Nilsson, Après ORIENT; A P&T-based New Resource Strategy in Nuclear Fuel Cycle, Global2013 Nuclear energy at a crossroads International Nuclear Fuel Cycle Conference, No.7477, 4x. Advanced Aqueous Separations, Salt Lake City, USA, September 29-October 3, 2013. 65
  • The Numerical Analysis about the Creation of Strategic Important Elements by Nuclear Transmutation Processes of Fission Products Atsunori Terashimaa*, Masaki Ozawaa a Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology *terashima.a.aa@m.titech.ac.jp Introduction The nuclear transmutation not only reduces unnecessary elements but can also create new elements. This study is aimed at verifying whether the nuclear transmutation by nuclear reactors has a possibility to change fission products (FP) into useful elements such as rare earth (RE) or platinum group metals (PGM). Calculation Method The condition of this study is shown in Fig.1 [1]. Firstly, the rates of transmutation from FP to some useful elements were calculated on the following condition. 1. Separation of FP from spent fuel. 2. Re-loading the FP element into light-water reactor (PWR) or fast breeder reactor (MONJU). 3. Irradiation of the FP element by excess neutrons in the reactor. Thus, correlation between the neutron capture cross-section and creation rates for each element was considered. Further, the isotopic composition of produced some useful elements and the time variation of these toxic of radioactivity were also calculated. In addition, if these radioactivity concentrations were higher than the exemption levels (E.L.) defined by IAEA [2], the cooling time which it took them to be less than the levels was calculated. Result and Discussion The calculation results of creation rates are shown in Fig.2. This figure shows that elements which are even atomic number (32Ge, 34Se, 36Kr, …) have higher creation rates than elements which are odd atomic number (31Ga, 33As, 35Br, …). This is based on the difference in the stability of the nucleus depending on the parity of the number of protons. That is, the nuclei with the even number of protons are more stable than those with the odd number of the protons. In addition, it turns out that there are two valleys (generally, 37Rb ~ 43Tc & 55Cs ~ 61Pm) which are lower creation rates inserted into three mountains which are higher creation rates. This reason is that there are neutron magic numbers in the valleys. In other words, some nuclei in the valleys are so stable that they are hard to cause neutron capture reaction. What is important information obtained from this figure is that many high creation rates are especially appeared in RE (for example, 64Gd : 23.4%/y (PWR), In the view 66Dy : 17.1%/y (MONJU)) and PGM (44Ru : 8.03%/y (MONJU), 46Pd : 20.6%/y (PWR)). of the short period of cooling, on the other hand, 39Y, 57La, 58Ce, 60Nd, 66Dy, 44Ru, and 46Pd are excellent product elements because they only require cooling time of 5 years or less. Conclusion On this calculation condition (Fig.1), As a result of comparing produced elements from some viewpoints, such as creation rates, radioactivity concentration, cooling time, and supply risk [3], 57La (from 56Ba), 60Nd (from 59Pr), 66Dy (from 65Tb), 44Ru (from 43Tc), and 46Pd (from 45Rh) can be considered as the important targets in future elements transmutation strategy. [1] Okumura Keisuke, Sugino Kazuteru, Kojima Kensuke, Jin Tomoyuki, Okamoto Tsutomu, and Katakura Junichi, “A Set of ORIGEN2 cross section libraries based on JENDL-4.0: ORLIBJ40,” JAEA-Data/Code 2012-032 (2013). [2] IAEA, “IAEA Safety Standards Series No.TS-R-1,” (1996). [3] U.S. Department of Energy, “U.S. Department of Energy Critical Materials Strategy,” (2010). 66
  • Calculation Code : ORIGEN2.2 Cross-section Library : ORLIBJ40 LWR PWR : Running (1125 days) , 45000MWd/tHM) (235U4.7% Cooling (5 years) Reprocess ing 30Zn ~ 69Tm Separating Elements from Fission Products Loading One Element with Fuel LWR Capture cross-section is large. Neutron Flux is large. FBR PWR : Running (1125 days) MONJU : Running (1125 days) (Neutron Flux : 3.29×1014/cm2/s) (Neutron Flux : 2.70×1015/cm2/s) Library : PWR47J40 Cooling (5 years) Library : MONJMXRDJ40 Fig.1. Calculation condition Fig.2. Comparison of creation rates of each element 67
  • Session 5: Hydro-Separation Technologies 68
  • Current Status of Reprocessing Process using Pyridine Resin in Hydrochloric Acid Solution Tatsuya Suzukia,*, Yu Tachibana a, Tomoo Yamamurab, Masanobu Nogamic, Shin-ichi Koyamad a Department of Nuclear System Safety Engineering, Nagaoka University of Technology b Institute for Materials Research, Tohoku University c Department of Electric and Electronic engineering, Kinki University c O-oarai Research and Development Center, Japan Atomic Energy Agency * Corresponding author: tasuzuki@vos.nagaokaut.ac.jp. Main Process We have been developing the reprocessing process using pyridine resin in hydrochloric acid solution. This reprocessing process is based on chromatographic techniques and considered for the partitioning for transmutation and utilization of valuable elements. In our presentation, the current status of the reprocessing process in hydrochloric acid solution and the standpoint from the partitioning for transmutation and utilization will be introduced. The schematic diagram of our proposing reprocessing process in hydrochloric acid (HCl) solution system is shown in Fig. 1. We have mainly used the pyridine resin as the separation materials. In the prefiltration and the main process, the pyridine resin is used. In the removal process of cesium and strontium, the separation material is not decided yet. This is one of remains. The chromatographic conditions in the prefiltration and the main process was decided by using the adsorption and separation results obtained by the batch experiments and the chromatography experiments. In the prefiltration and the main process, we carried out the experimental tests by using the irradiated MOX fuel solution and the simulated spent fuel solutions. From results of these experiments, the behavior of many fission product elements and actinides in chromatography was understood. The FP(Sr, Cs, Ln, etc), PGM, Spent Fuel U, Np, Pu, MA(III) remained important elements is neptunium. We investigated the adsorption behavior of Plasma Voloxidation Dissolution Process neptunium ions on pyridine resin in HCl UO2 g U3O8 solution[1]. The valence of neptunium was Dissolution by HCl+H2O2 STEP I Removal of Cs, Sr controlled to the tetravalent, the pentavalent and the hexavalent. These valence was STEP II PGM, Tc, etc. Pre-filtration confirmed by UV-Vis. spectrometry. The Removal & Recovery 0.5M HCl obtained distribution coefficients are shown FP(Ln, etc.), in Fig. 2. The adsorptions of the pentavalent U, NP, Pu, MA(III) STEP III neptunium ion under 3.5 mol/L of HCl and 12M HCl U, Np&Pu of the,hexavalent neptunium ion under 1 mol/L of HCl were not observed. The 0.5M HCl MA(III) distribution coefficients of the pentavalent Fp(Ln, etc.) neptunium ion over 11.5 mol/L of HCl were 0.5M HCl U, Np&Pu not obtained because of strongly absorbing All An MA(III) on pyridine resin. The adsorption behavior FP(Ln, etc.) 12M HCl of the valence controlled neptunium ions Fig. 1. Schematic diagram of our proposing need be investigated more in detail. However, we knew from these results that reprocessing process in hydrochloric acid solution. any valence of neptunium ions are adsorbed 69
  • Distirbution coefficient on pyridine resin under the condition of main process. This fact shows that neptunium is Np(VI) recovered with the uranium and the plutonium. One of important remains is dissolution process. 2 10 We propose the plasma voloxidation and the Np(V) dissolution in HCl solution by adding hydrogen peroxide (H 2 O 2 ). The oxygen plasma can transform from UO 2 pellet to U 3 O 8 powder under Np(IV) the lower temperature, i.e., around the room 0 temperature. The oxidation and the powderization 10 of uranium oxide make the dissolution of uranium oxide in HCl accelerate. The addition of H 2 O 2 make the oxidation environment in HCl solution, 0 4 8 12 and by this oxidation environment the dissolution HCl / mol/L is promoted. It was clarified that our proposing Fig. 2. Distribution coefficients of neptunium in reprocessing process can recover all actinides, hydrochloric acid solution. especially uranium, neptunium, and plutonium are recovered simultaneously. The non-isolation process of plutonium is important from the view of nonproliferation. We believe that our proposing reprocessing process is available to be used for the future reprocessing process. Recently, ADS (accelerator driven system, or accelerator driven subcritical reactor system) is attracted. ADS is planned to use the nitride fuel made of nitrogen-15. Thus, the recovery of nitrogen-15 in the reprocessing process is required, i.e. the nitric acid cannot used for the dissociation. Our processing process is available for ADS reprocessing process, because our process uses the hydrochloric acid instead of the nitric acid. This work was partially supported by the Grant-in-Aid for Scientific Research (B) (KAKENHI No. 23360423). [1] Y. Tachibana, et al. submitted to J. Radioanal. Nucl. Chem. 70
  • Studies on the Advanced Hybrid Reprocessing System “FluoMato” Process Yuezhou Wei a,b*, Ruiqin Liu a, Yan Wu a, Long Zhao a, Jianhua Zu a, Hitoshi Mimura b, Weiqun Shi c, Zhifang Chai c, a Shanghai Jiao Tong University, 800 Dongchuan Rd., Shanghai 200240, China 1 b Tohoku University, Aoba-ku, Sendai 980-8578 c Institute of High Energy Physics, Chinese Academy of Science, Beijing 100049, China *Corresponding author: yzwei@sjtu.edu.cn Although the Purex process is a relatively matured technology and has been applied to commercial plants for reprocessing spent LWR-fuels, there are also some significant drawbacks: (1) It uses a great amount of organic solvent/diluent which results in a large scale equipment and a large amount of waste generation, (2) It can recover U and Pu only, and (3) It is hardly applicable to future high-Pu containing fuels such as LWR-MOX fuel and FBR fuel. Fig.1 gives the composition examples of different type of spent nuclear fuels. The content of Pu in spent FBR-MOX fuel is about 10 times of that in a common spent LWR fuel. In the conventional PUREX process treating LWR-fuel, generally more than 5 times of stoichiometric quantity of a reducing agent such as U(IV) is used to reduce the extractable Pu(IV) to inextractable Pu(III) for the partitioning of Pu from U. If such reductive stripping technique is applied to the high fractional Pu(IV) reduction, a significantly large amount of reducing agent must be required and the feasibility of this method is very questionable. In addition, the content of minor actinides (MA: Np, Am, Cm), especially Am and Cm in spent LWR-MOX and FBR fuels is 5-10 times higher than that in LWR-fuel. These long-lived MA nuclides will give significant impact on the final disposal of high level waste1). In recent years, to recover U, Pu, MA and some specific FPs (Cs, Sr, Tc, etc.) from different type of spent nuclear fuels including LWR-UO2 fuel, LWR-MOX fuel, FBR MOX and metal fuels, we are studying an advanced hybrid reprocessing system named “FluoMato” process, which is based on UF6 volatilization (Pyro) and chromatographic separation (Aqueous) methods. The schematic flow sheet of the FluoMato process is shown in Fig.2. The pyro-process is similar to that in the FLUOREX process developed by Hitachi-GE2). According to their experiment results using simulated spent fuel, more than 90% of U was fluorinated to volatile UF6 and then recovered. After the recovery of the most amount of U, the mass of residual fuel components which will be treated in the aqueous process is significantly reduced, since Pu, MA and FPs are much less abundant than U in spent fuels. So the scale of the subsequent aqueous separation process could become reasonably small and result in less waste. For the chromatographic separation processes (aqueous), we have prepared different types of porous silica-based organic/inorganic adsorbents with fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. So they are advantageously applicable to efficient separation of the actinides and FP elements from the residual fuel dissolved solution. Fig.3 shows the chromatographic separation process which mainly consists of the following separation steps: (1) Cs and Sr are selectively adsorbed and separated by using porous SiO2 based salt adsorbents containing ammonium molybdophosphate, ferrocyanide or titanate. Selective separation (removal) of 137Cs and 99Sr from the residual fuel dissolved solution at early stage is advantageous to diminish radiation damage to organic absorbents used in the subsequent processes. (2) Pu, Np and residual U can be efficiently recovered from the remained fuel dissolved solution by anion exchange column packed with the novel silica-based AR-01 anion resin, since they form anionic nitratocomplexes in a concentrated HNO3 solution3). Moreover, Tc strongly adsorbed onto AR-01 as the form of TcO4- and also can be isolated by the AR-01 column. (3) The MA (Am, Cm) separation process consists of two chromatographic columns using novel silica-based adsorbents containing CMPO or TODGA and R-BTP, respectively. In the CMPO (or TODGA) column, the elements contained in fuel solution containing concentrated HNO3 can be partitioned into (a) Non-adsorptive FP, (b) MA-Ln 71
  • (lanthanides), and (c) Zr-Mo, depending on their different adsorption-elution behavior. The effluent of MA-Ln is then applied to the second column packed with an R-BTP adsorbent, where the MA can be separated from the Ln, since R-BTP shows high adsorption selectivity for MA over Ln4). In this work, adsorption and separation behavior of representative actinides and FP elements was studied experimentally. Small scale separation tests using simulated and genuine fuel dissolved solutions were carried out to verify the feasibility of the proposed aqueous process. [1] J. Magill, V. Berthou, D. Haas, Nucl. Energy. 42, 263 (2003). [2] K. Hoshino, D. Watanabe, A. Sasahira, M. Nagata, T. Chikazawa, Y.Sano, F. Kawamura, Proc. GLOBAL 2011, Makuhari, P.No.391103(2011). [3] Y.-Z. Wei, T. Arai, H. Hoshi, M. Kumagai, A. Bruggeman and P. Goethals,Nucl. Technol., 149, 217 (2005). [4] Y.-Z. Wei, H. Hoshi, Y.Morita, A. Bruggeman, and P. Goethals, Proc. Global 2009, Paris, Sep.6-11, 2009, Paper 9390 (2009). FP, MA(5%) Pu(1%) U(94%) Accumulation of MA Spent LWR Fuel Fuel Solution (3M HNO3) Pu(10%) FP, MA(8%) Inorganic salts /SiO2 U(82%) FBR-MOX Fuel (an example) Cs/Sr adsorbent column Other Elements Anion exchanger column Cs/Sr AR-01/SiO2 Cooling time MA-FPs U/Pu/Np Tc Fig.1 Composition of different types of spent fuel Chelate (CMPO, TODGA, R-BTP) adsorbent /SiO2-P column Other FPs Am/Cm Pd Fig.3 The aqueous process based on chromatographic separation Fig. 2 Schematic flowsheet of the FluoMato Process 72
  • R&D Efforts Using Novel Extractants for the Development of ‘Green’ Separation Technologies Relevant in the Back-End of Nuclear Fuel Cycle P.K. Mohapatra Radiochemistry Division, Bhabha Atomic Research Centre, Trombay, Mumbai – 400 085, INDIA E-mail: mpatra@barc.gov.in With the ever increasing demand for energy and limited availability of fossil fuels such as oil and coal, non-conventional energy resources such as renewable and nuclear energy are going meet most of the demand. In the nuclear energy sector, need for the fast breeder reactor operation requires countries to opt for the closed fuel cycle. In India, in view of the huge thorium reserves, there is a great deal of effort being made in developing thorium based reactors. Furthermore, the need for reducing the long term hazards of the radiotoxic fission product and actinide elements has led to a great deal of R&D efforts being directed in the area of radioactive waste management. This provides major motivation to develop novel schemes for the separation of uranium, and plutonium from other elements with high decontamination factors (DFs) and also to develop novel separation methods for the recovery of minor actinides and long-lived fission products from the radioactive wastes. The challenging task of recovery and purification of 239Pu from irradiated U and of 233U from irradiated Th are conventionally carried out by the versatile PUREX and THOREX processes, respectively [1]. However, the experience gained over last five decades on the reprocessing of spent fuel has identified some major drawbacks of TBP which result in deterioration of process efficiency and generation of large volumes of non-incinerable wastes. These shortcomings are of particular concern during the during the reprocessing of short-cooled (MOX) thermal reactor, Advanced Heavy Water Reactor (AHWR, being developed in India based on Th) as well as fast reactor spent fuels with larger Pu content and significantly higher burn up [2]. In this context, N,N-dialkyl amides appear attractive option as potential extracting agents for actinides during the reprocessing of irradiated fuels. Extensive studies carried out at Radiochemistry Division, BARC on the evaluation of N,N-dialkyl amides as extractants in the back-end of the nuclear fuel cycle for addressing the issues related to the reprocessing of U and Th based spent fuels. Amongst these amides, N,N-dihexyloctanamide (DHOA) and N,N-di(2-ethylhexyl)isobutyramide (D2EHIBA) received considerable attention as potential alternatives extractants for the reprocessing of U and Th based spent nuclear fuels, respectively [3]. Batch extraction, mixer settler/ centrifugal contactor runs were carried out with these reagents under Pu rich feed (relevant to fast reactor) and AHWR feed conditions, respectively. Solvent (0.05M TODGA + 5% Isodecanol / dd) Scrub (5M HNO3 + 0.2M oxalic acid + 0.1M HEDTA) Extraction Stages 1 - 4 – 8 - 12 Raffinate Scrubbing Stages 13 – 17 – 21 – 24 Feed (0.1M oxalic acid+ 0.05M HEDTA) Spent solvent Stripping stages 25 – 28 – 32 - 36 Product (Ln/An) Strip (0.2M HNO3) Organic phase Aqueous phase Fig.1: Proposed flow sheet for the partitioning of minor actinides employing TODGA as the extractant under simulated Pressurized Heavy Water Reactor –HLW conditions The Partitioning & Transmutation (P&T) strategy envisages the quantitative removal of minor actinides from radioactive waste and their subsequent burning in high flux reactors / accelerators in suitable chemical forms. Exercise of P&T option seeks to convert all long-lived nuclides into shortlived (or stable) species. This process apart from generation of energy alleviates the need for the longterm surveillance of geological repositories. In this context, several extractants such as Octyl(phenyl)-N,N-diisobutyl carbamoyl methyl phosphine oxide (CMPO), N,N’-dimethyl-N,N’-dibutyl tetradecyl malonamide (DMDBTDMA), N,N’-dimethyl-N,N’-dioctyl hexylethoxy malonamide 73
  • (DMDOHEMA), N,N,N′,N′-tetraoctyl diglycolamide (TODGA) and N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (TEHDGA) have been systematically evaluated (by batch as well as counter-current mixer settler runs) for the partitioning of minor actinides from high level waste (HLW) solutions [4,5]. TODGA based process flow sheet has been developed for lab scale actinide partitioning studies using simulated feeds (Fig. 1). In addition, diglycolamide-functionalized calix[4]arenes (DGA-Calix) were also evaluated for actinide extraction from acidic feeds. The ligands with four diglycolamide (DGA) pendent arms are significantly more effective extractants than those with two DGA pendent arms. The ligands showed a preference for the extraction of Eu3+, a representative trivalent lanthanide ion, as compared to Am3+, a commonly encountered trivalent actinide ion [6]. The group separation of Ln(III)-An(III) is a challenge prior to the transmutation step in view of the chemical similarity of these metal ions. It requires special efforts to identify suitable extractants (N,S donors) to achieve the desired separation. Extraction studies carried out in our laboratory on using this class of ligands have shown very promising results for future scale up studies [7]. The recovery of heat emitting radio-cesium and radio-strontium from HLW solutions is a major challenge in the back-end of fuel cycle. This has a bearing on the stability of the vitrified glass matrix for disposal in deep geological repositories. Efforts have been made to optimize the conditions for their recovery from HLW solutions using crown ethers and their calix-arene derivatives [8]. These studies include development of the extraction as well as membrane based separation techniques. This presentation will summarize all the separation studies on the evaluation of novel ligands relevant to the back-end of fuel cycle. References 1. W.W. Schulz, L.L. Burger, J.D. Navratil, K.P. Bender (Eds.) (1990) Science and Technology of Tributyl phosphate, Vol. III, CRC Press, Inc. Boca Raton, Florida. 2. R.K. Sinha, A. Kakodkar, (2006) Nucl. Eng. Design, 236 (2006) 683. 3. V.K. Manchanda, P.N. Pathak, Sep. & Purif. Technol., 35 (2004) 85. 4. S.A. Ansari, P.N. Pathak, P.K. Mohapatra, V.K. Manchanda. Chemical Reviews, 112 (2012) 1751. 5. S.A. Ansari, P.N. Pathak, P.K. Mohapatra, V.K. Manchanda. Sep. & Purif. Reviews, 40 (2011) 43. 6. D.R. Raut, P.K. Mohapatra, S.A. Ansari, S.V. Godbole, M. Iqbal, D. Manna, T.K. Ghanty, J. Huskens, W. Verboom, RSC Adv., 2013,3, 9296-9303 7. A. Bhattacharyya, P.K. Mohapatra, V.K. Manchanda, Solv. Extr. Ion Exch., 24 (2006)1. 8. D.R. Raut, P.K. Mohapatra, V.K. Manchanda, Sep. Sci. Technol., 45(2010)204. 74
  • Preparation of high purity thorium by Centrifugal extraction a Yanliang Wanga, Yanling Li, a Changjun Zhao,a Guolong Wua, Wuping Liaoa,*, Deqian Lia State Key Laboratory of Rare Earth Resource Utilization, Changchun Institute of Applied Chemistry, Chinese Academy of Sciences, Changchun 130022, China * Corresponding author: wpliao@ciac.ac.cn Thorium is considered to be an important nuclear fuel because 232Th is convertible to fissile 233U by absorbing slow neutrons. Commonl y, thorium associates with rare earths in different minerals such as monazite and bastnasite. In China, Baotou deposit and Sichuan deposit, which have about 95% rare earth resource of China, also contain about 240,000 ton of thorium. In the past decades, we developed a solvent extraction process to recover thorium from the sulfate leaching of these two rare earth deposits using primary amine extractant N1923. Most thorium can be recovered from the leaching and the purity of thorium product reaches 99%. Aiming to its use in nuclear energy, the raw thorium product was further purified by the solvent extraction with a novel extractant DEHEHP. It is found that DEHEHP exhibits good extraction ability towards thorium and thorium can be stripped easily from the loaded organic phase by water. By a multistage fractional extraction using centrifugal extractors, the purity of thorium reaches 99.99%. After another multistage centrifugal extraction, the thorium purity is increased to 99.999%. Several kilogram of high purity thorium was obtained by this process. [1] Y. L. Wang, Y. L. Li, W. P. Liao and D. Q. Li, J. Radioanal. Nucl. Chem. DOI: 10.1007/s10967-013-2643-3(2013). [2] Y. L. Wang, Y. L. Li, D. Q. Li and W. P. Liao, Hydrometallurgy DOI: 10.1016/j.hydromet. 2013.09. 002(2013). [3] D. Q. Li, Y. L. Wang and W. P. Liao, China patent, ZL ZL201110074345.8. [4] W. P. Liao, Y. L. Wang, D. Q. Li and C. J. Zhao, China patent, appl. no.201310105102.5. Fig. 1 Effect of acid concentration on the extraction (left) and the stripping isotherms (right). Fig. 2 Process scheme for thorium purification by a multistage centrifugal extraction. 75
  • Abstract for ASNFC 2013 Development of Selective Separation Method for Nuclear Rare Metals using Highly Functional Xerogel Microcapsules Hitoshi Mimuraa, Takashi Ohnishib, Hiroshi Ohbayashib, Shin-ichi Koyamab, Rana Syed Masuda, Takuya Kawamuraa, Yuichi Niiboria, Hiroshi Sugaib a Tohoku University, Japan b Japan Atomic Energy Agency, Japan c 3R Co. Ltd., Japan Special attention has been given to the selective separation and recovery of heat-generating nuclide (Cs), platinum group metals (PGMs: Pd, Ru and Rh) and oxoanions (Mo, Re and Tc, etc) from high-level liquid wastes (HLLWs) in relation to the partitioning of radioactive nuclides and their effective utilization. As for the recovery of these nuclides from HLLWs, selective inorganic ion-exchangers and extractants have been developed, and their granulation with alginate biopolymers is effective for the compact and successive separation process using the packed column. We have attempted to immobilize these materials into the porous alginate gel polymers; the microcapsules (MCs) can be readily prepared by using sol-gel method. The present study deals with the uptake behavior of the above nuclides on MCs, characterization and the chromatographic separation for simulated HLLW (28 components, SW-11E, JAEA) and real waste (dissolved waste solution from spent MOX, JAEA) by batch and column methods. Granular MCs of AWP (ammonium tungstophosphate) encapsulated with calcium alginate (CaALG) matrices exhibited a high selectivity towards Cs+, and a relatively large separation factor (α) of Cs/Rb above 100 was obtained. The mutual separation of Cs/Rb was thus accomplished through the packed column and the resolution of Cs/Rb by the chromatography was estimated to be 1.0. The novel MCs of insoluble ferrocyanides (KCuFC) prepared by direct method had an excellent selectivity towards Pd2+; relatively large K d values above 105 cm3/g were obtained. The chromatographic separation of Pd/Rh/Ru in simulated HLLW was accomplished using thiourea-HNO 3 eluent. The extractants (LIX63, MIDOA and TOA) can be encapsulated with the polymer gels of Ca-H-ALG. The MCs enclosing the above extractants exhibited a high selectivity towards Mo, Re and Tc, and the chromatographic separation for simulated and real wastes solution was accomplished by using stepwise elution methods (Fig. 1). The precise separation system combined the above processes is effective for the recovery of nuclear rare metals from HLLWs. Relative concentration / - H2O 1M NH4NO3 1 0.8 0.6 0.4 0.2 0 0 10 Fig. 1 20 Volume / cm3 30 40 Rb Y Zr Tc Ru Rh Pd Ag Te Cs La Ce Pr Nd Sm Eu 80 H2O 70 2M HNO3 5M HNO3 60 Elution / % 1.2 50 40 30 20 10 0 0 10 20 30 40 50 60 70 Volume / cm3 Chromatographic separation of Cs (left) and Mo (right) for real HLLW (JAEA). 76 Y Mo Tc Ru Rh Te Cs La Ce Pr Nd Sm Eu Gd
  • Novel Pillar[5]arene-Based Phosphine Oxides as Extractants for the Segregation of f-Block Elements from Acidic Media in Biphasic Systems Lihua Yuan, Yuyu Fang, Lei Wu, Long Chen, Jiali Liao, Yuanyou Yang, Ning Liu, Wen Feng* Key Laboratory for Radiation Physics and Technology of Ministry of Education, Institute of Nuclear Science and Technology, College of Chemistry, Sichuan University, Chengdu, 610064, P. R. China * Corresponding author: wfeng9510@scu.edu.cn. Abstract: Three pillar[5]arene-based phosphine oxides tethered with ten chelating groups on both sides of the pillar were synthesized. These macrocycles were found to demonstrate high efficiency and selectivity in extraction of some representative lanthanides and actinides, especially Th(IV) and U(VI) in acidic media. In addition, the moderate efficiency was observed in differentiating europirm(III) and americium(III) in the presence of a synergist (Br6-COSAN). The efficient differentiation of actinides (An) from lanthanides (Ln) is one of the most important issues both in the process of hydrometallurgy and management of nuclear fuel [1]. Solvent extraction is a widely-investigated reprocessing method exploited for f-elements and other cations discrimination due to its simplicity and easy operation [2].The obtained extraction and separation ability is mainly dependent on the chemical structure of the extractants. Preorganizing multiple chelating groups onto a macrocyclic platform or rigid scaffold pertinent to entropic stabilization results in increasing complexation constants, separation efficiency and selectivity [ 3 ]. Pillararenes are a new class of macrocyclic compounds with a highly symmetrical pillar architecture that differs from the typical calixarenes in basket conformation [4]. They have received great and continuous attention since their discovery [5]. It is expected that these macrocycles should be suitable to serve as a platform for preorganizing chelating groups for metal ion separation. Herein, we designed and synthesized a new class of pillar[5]arene-based ligands 1a-c functionalized with phosphine oxide groups (Fig.1) and demonstrated its capability in extraction towards selective separation of some representative lanthanides and actinides. The target molecules and intermediates were characterized by NMR and HR-ESI-MS. n Ph O P Ph n Ph O P Ph O O O P CH2 O O n n 5 TOPO Ph P O Ph 1 a, n = 0 b, n = 1 c, n = 2 Ph P O Ph 2 Fig.1 The chemical structures of pillar[5]arene-based phosphine oxides 1, 2 and TOPO. Solvent extraction of the ligands 1 towards selected lanthanides and two actinide cations Th4+ and UO22+ was investigated. The remarkable efficiency and selectivity for Th4+ and UO22+ were observed compared to that for their corresponding anolgues 2 (Fig.2a). With increasing spacer length in the case of 1, the extraction efficiency was greatly enhanced from 77
  • 27% to 72% for Th4+, and from 61% to 94% for UO22+, with 1a being the least efficient. As shown in Fig 2b, with increasing acidity, the extraction for Th4+ and UO22+ using 1b resulted in increment of D values, while this trend was reversed for TOPO at the concentration that is 10-fold (UO22+) or 25-fold (Th4+) higher than that of 1b. Fig.2. (a) Extraction percentage (%E) of f-element nitrates (10-4 M) from 1 M HNO3 aqueous solution into dichloromethane containing the ligands 1 (10-3 M); (b) Influence of the nitric acid concentration in the aqueous phase on the extraction of Th4+ and UO22+ by 1b and TOPO from aqueous solution into dichloromethane ([1b]=10-4 M for UO22+; [1b]=10-3 M for Th4+; [TOPO]=10-2 M for UO22+; [TOPO]=0.025 M for Th4+, T=20°C). The extraction of Th4+ and UO22+ at various sodium nitrate concentrations was also investigated. The extractability increased with increasing NaNO3 concentration in the presence of 1 M HNO3 (Fig.3a). The higher the concentration of NaNO3, the better the extraction for selectively separating U(VI) and Th(IV). The stoichiometry of the extracted Th4+ and UO22+ complexes with 1b indicated that the ligand may extract the cation as 1:1 species (Fig.3b). The moderate separation capability for Eu3+ versus Am3+ of these ligands was observed in the presence of a synergist (Br6-COSAN) under specified acidic condition. Fig.3. (a) Extraction percentage (%E) and separation factor (SF) of Th4+ and UO22+ from 1 M HNO3 and various sodium nitrate aqueous solution into dichloromethane containing the ligand 1b (10-4 M); (b) Plots of Log DM versus Log CL for the extraction of Th4+ and UO22+ by 1b and TOPO from 1 M HNO3 aqueous solution into dichloromethane (T=20°C). Acknowledgement This work is supported by NSAF(11076018) and Sichuan Province Science and Technology Support Programme (2011FZ0048) [1] Z. Kolarik, Chem. Rev. 108, 4208 (2008). [2] H. Eccles, Solvent Extr. Ion Exch. 18, 633 (2000). [3] H. H. Dam, D. N. Reinhoudt and W. Verboom, Chem. Soc. Rev. 36, 367 (2007). [4] T. Ogoshi, S. Kanai, S. Fujinami, T. Yamagishi and Y. Nakamoto, J. Am. Chem. Soc. 130, 5022 (2008). [5] M. Xue, Y. Yang, X. Chi, Z. Zhang and F. Huang, Acc. Chem. Res. 45, 1294 (2012). 78
  • Synthesis and adsorptivity of acryloylmorpholine resin for selective separation of U(VI) in nitric acid media Tetsuhiro NISHIDA, Shotaro TANAKA, Masanobu NOGAMI* Department of Electric and Electronic Engineering, Faculty of Science and Engineering, Kinki University * Corresponding author: mnogami@ele.kindai.ac.jp Uranium(VI) is the most stable U species in aqueous nitric acid solutions. Separation of U(VI) from HNO 3 containing U(VI) and other metal ions is very important to treat radioactive wastes. For developing resins with selectivity to U(VI) in HNO 3 media, we have synthesized several silica-supported polymer beads with the structure of a monoamide as the functional group and their adsorptivities to various metal ions have been examined[1-4]. In addition, we have investigated the adsorptivity of commercially-available polyvinylpolypyrrolidone (PVPP), one of cyclic monoamide resins with a 5-membered pyrrolidone ring (Fig. 1.(a))[5]. Adsorptivity of these resins to metal ions except U(VI) has been found different. Namely, PVPP shows weak adsorptions for Pd(II) and Re(VII) in HNO 3 of lower concentration range. On the other hand, a silica-supported resin consisting of N,N-dimethylacrylamide (Silica-DMAA), a chain monoamide (Fig. 1.(b)), exhibits no adsorption for major metal ions of fission product elements regardless of the concentration of HNO 3 . In the present study, a novel silica-supported resin with a monoamide of 4-acryloylmorpholine (Silica-AM) (Fig. 1.(c)) was synthesized, and the adsorptivity to metal ions (a) (b) (c) except U(VI) was examined. AM has some interesting features from the viewpoint of the chemical structure; (i) it has a 6-membered ring, (ii) the carbonyl group is not included in the ring unlike PVPP, (iii) the ring has a coordinative ether oxygen atom. It was, therefore, expected that the adsorptivity of Silica-AM is more Fig. 1. Chemical structure of resins: complicated than those of PVPP and DMAA. (a) PVPP (b) DMAA (c) AM Silica-AM was synthesized in a similar procedure to the synthesis of Silica-DMAA[1]. As the results of the batch adsorption experiments for the obtained Silica-AM, it was found that Pd(II) and Re(VII) were adsorbed in HNO 3 of lower concentration range like PVPP, and that the dependence of the adsorptivities on HNO 3 concentration a little differed from that of PVPP. It was also found that Mo(VI) and Zr(IV) were adsorbed in HNO 3 of lower concentration range unlike PVPP. Considering the difference in the chemical structure of the resins, oxygen atoms of the carbonyl group and ether may contribute mainly to the adsorption of Pd(II) and Re(VII), and Mo(VI) and Zr(IV), respectively. [1] M. Nogami, T. Ishihara, K. Suzuki and Y. Ikeda, J. Radioanal. Nucl. Chem., 273, 37 (2007) [2] M. Nogami, T. Ishihara, K. Maruyama, and Y. Ikeda, Prog. Nucl. Energy, 50, 462 (2008) [3] M. Nogami, Y. Sugiyama, and Y. Ikeda, J. Radioanal. Nucl. Chem., 283, 177 (2010) [4] M. Nogami, T. Nishida, and N. Miyata, Energy Procedia, 39, 96 (2013) [5] M. Nogami Y. Sugiyama, T. Kawasaki, M. Harada, Y. Morita, T. Kikuchi, Y. Ike da, J. Radioanal. Nucl. Chem., 283, 541 (2010) 79
  • Adsorption behavior of Am(III) and Ln(III) from Nitric Acid Solution onto isoHexylBTP/SiO 2 -P Adsorbent Ruiqin Liua, Xinpeng Wanga, Shunyan Ninga, Jianhua Zua, Yuezhou Weia*, Youqian Dingb, Jinling Yang b, Weiqun Shic, Zhifang Chaic a School of Nuclear Science and Engineering, Shanghai Jiaotong University, 800 Dongchuan Road, Shanghai, 200240, China b China Institute of Atomic Energy, Beijing, 102413, China c Institute of High Energy Physics, Chinese Academy of Science, Beijing 100049, China * Corresponding author: yzwei@sjtu.edu.cn In order to develop a separation process for trivalent minor actinides from fission products in high level liquid waste (HLLW) by extraction chromatography, a macro porous silica-based adsorbent isoHexyl-BTP/SiO 2 -P was prepared. The adsorbent exhibited much higher adsorption affinity for Am(III) in 2-3 mol/L HNO 3 solution over fission products (including Ln(III)) which are contained in HLLW. The effects of adsorption parameters of the nitric acid concentration of solution, contact time, initial Dy(III) concentration and temperature for Dy(III) adsorption onto the isoHexyl-BTP/SiO 2 -P adsorbent were investigated. The obtained adsorption data depending on contact time were analyzed by using adsorption models such as the pseudo-second-order, the particle diffusion, and the intra-particle pore diffusion models. The adsorption data obtained from experimental results depending on the equilibrium solution concentration were analyzed by the Freundlich, Langmuir adsorption isotherms. The obtained experimental adsorption data depending on temperature were evaluated to calculate the thermodynamic parameters of enthalpy (∆H), entropy (∆S) and free energy change (∆G) for the Dy(III) adsorption onto the isoHexyl-BTP/SiO 2 -P adsorbent from nitric acid solutions. [1] Y.Z. Wei, A. Y. Zhang, M. Kumagai, M. Watanabe, N. Hayashi, J. Nucl. Sci. Technol. 41(3), 315 (2004). [2] Y. Z. Wei, M. Kumagai, Y. Takashima, G. Modolo, R. Odoj, Nucl. Technol. 132(3), 413 (2000). [3] R. Q. Liu, Y. Z. Wei, D. Tozawa, Y. L. Xu, S. Usuda, H. Yamazaki, K. Ishii, Y. Sano, Y.Koma, Nucl. Sci. Techn. 22(1), 18 (2011). [4] H. Hoshi, Y.Z. Wei, M. Kumagai, T. Asakura, Y. Morita, J. Alloys Compd. 408-412, 1274(2006). [5] Y. Z. Wei, H.Hoshi, M. Kumagai, T. Asakurab, Y. Moritab, J. Alloys Compd. 374, 447 (2004). 2.0x104 4 152 Am Eu 1.6x10 0.15 1.2x104 Qt (mmol/g) kd (cm3/g) 0.18 241 8.0x103 4.0x103 0.0 0 1 2 3 HNO3 concentration (mol/L) 0.12 0.09 0.06 0.03 0.00 0 4 Fig. 1. Element distribution coefficients of 241 Am(III) and 152Eu(III) as a function of the initial HNO3 concentration (298 K, phase tracer (241Am(III), ratio: 0.1g/5cm3, 152 Eu(III)), contact time: 24 h). 288 K 298 K 308 K 20 40 t (h) 60 80 Fig. 2. Effect of contact time on Q of Dy(III) onto isoHexyl-BTP/SiO2-P adsorbent from solution containing 6 mM Dy(III) and 3 M HNO3 (phase ratio 0.1 g/5 cm3) 80
  • Preparation of anion exchanger by pre-irradiation grafting method and its adsorptive removal of Rhenium as an analogue of radioactive technetium Jianhua Zua * Ruiqin Liu a Yuezhou Wei a Fangdong Tangb Linfeng Heb a School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai b Shanghai Institute of Measurement and Testing Technology, Shanghai * Corresponding author: zujianhua@sjtu.edu.cn. 99 Tc is a long-lived, β-emitting fission product with a half-life of 2×105 years and a relatively high fission yield (6.3% of 235U). Removal of this isotope from spent nuclear fuel waste streams is important in view of its long lifetime, toxicity and high mobility in the environment. Rhenium, which is placed in the VIIB group of the periodic table together with Tc, is a good chemical analogue of Tc owing to their similar electronic configuration and stereochemistry and thermodynamic properties. So the related research results of this paper will provide very important experimental data for separation of technetium. Kim et al demonstrated that ReO4- which is often used as pertechnetate analogue, is efficiently absorbed by natural organic polymer chitosan containing amino groups [1]. Liang et al showed that adsorption of pertechnetate by Forager sponge can also be attributed to the presence of amino groups [2]. Therefore, on the basis of our previous extensive research as well as on the literature data, we concluded that amino-functionalized copolymers may be used for pertechnetate adsorption. In this article, to develop an advanced ion exchanger which has high stability and uses compacted equipment for separation of Tc, a new anion exchanger with a quaternary ammonium(-N+(C2H5)3) functional groups was prepared by grafting of 2vinyl pyridine (2-VP) onto polypropylene (PP) non-woven fabrics using pre-irradiation technology and subsequent quaternization of graft copolymers by reacting with bromoethane. Results showed that the grafting yield increased with the increase of monomer concentrations and the longer quaternization time gained the higher conversion ratio. The grafted and quaternized nonwoven fabrics were characterized by FT-IR and DSC. The possibility of adsorption of Perrhenate (ReO4-), a nonradioactive surrogate for pertechnetate (99TcO4-) from aqueous solution by anion exchanger was investigated. The results of batch experiments performed at pH 0.1-6 showed pH2.2 was the optimal acidity for adsorption of ReO4- and the absorption equilibrium achieved within 30min. ∆H was -12.55kJ/mol which indicated that the adsorption process was exothermic reaction. XPS tests indicated the nature of Rhenium uptake was typical ion exchange between Cl- on anion exchanger and ReO4-. [1] E. Kim, M. Benedetti, F. Boulegue, J .Water Res. 448(2004). [2] L. Ling, B. Gu, X. Yin, Sepa. Tech. 111(1996). Intensity 40000 20000 0 1200 1000 800 600 400 200 0 Binding Energy (eV) Preparation of anion exchanger by radiation induced grafting and quaternization XPS spectra of anion exchanger with adsorption of Re(Ⅶ) 81
  • Adsorption of Th4+ from aqueous solution onto Poly(N,N-diethylacrylamide-co-acrylic acid) Microgels Liu Tonghuan*, Hu Peizhuo, Duan Xiaojiang, Wu Wangsuo Radiochemistry Laboratory, School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000, China Thorium is used as a fertile material for producing nuclear fuel. Since thorium is more abundant than uranium, some nuclear reactor designs incorporate thorium in their fuel cycle. It may also be used directly as nuclear fuel instead of uranium, producing less transuranic waste [1]. In the industry, Th is used as an alloying element in magnesium, used in manufacture of aircraft engines, imparting high strength and creep resistance at elevated temperatures and to coat tungsten wire utilized in electronic equipment, improving the electron emission of heated cathodes [2]. Beside its industrial importance, Th ions are extremely dangerous for the environment and human health such as lung, pancreatic and liver cancer by their high toxicity even at very low concentrations and long half lives [3] . Both industrial importance and toxicity signify the need of Th removal and recovery. The methods developed for the removal of radioisotopes from the wastewaters include: chemical precipitation, ion exchange, membrane-related processes, biological processes and electrochemical techniques[4-6] However, excessive time requirements, high costs, and production of highly toxic sludge are the constraints of these techniques,[7]and other novel techniques have been developed.[8] The adsorption procedures mainly involve the development of low cost adsorbents with easy manipulation and regeneration for possible reuse[9,10]. with recent developments in resin technology, a great deal of attention has been paid to polymer materials due to their biodegradability and non-toxic nature, [11, 12] which have suitable functional groups (such as carboxylate) containing O, N, as well as S and P as donor atoms capable of interaction with metal ions.[13] In previous work, we have reported the synthesis of Poly(N, N-diethylacrylamide-co-acryliIcacid) (P(DEA-co-AA)) microgels and their application [14]. As part of our continuing investigation into the development of microgels involving metal ions, herein cross-linked (P(DEA-co-AA)) microgels was synthesized by the reaction of radical polymerization, which cannot be dissolved in the aqueous solution and organic solvents. This adsorbent is expected to show higher affinity to metal ions via the chelating interaction between 82
  • the Th4+. The effects of contact time, solid/liquid ratio, pH value, and initial concentration on the adsorption of Th4+ ions onto P(DEA-co-AA) were investigated. The adsorption of Th4+ ions was highly dependent on the initial pH of metal ions solution and initial metal ions concentration. The adsorption kinetic data indicated that the chemical adsorption was the swiftness processes, the adsorption equilibrium could be achieved within 20 min. And there are very good correlation coefficients of linearized equations for Langmuir isotherm model, which indicated that the sorption isotherm of the hydrogel for Th4+ can be fitted to the Langmuir isotherm model. Gibbs free enthalpy change was ∆G< 0 indicating that the adsorption process is spontaneous. The enthalpy and entropy changes were ∆H◦>0 and∆ S◦>0 for both adsorbents and for both ions showing that the overall process was endothermic. After five times of repeated tests for the PDEA-co-AA microgels it still remained its excellent adsorption. Fig. 1. The amount adsorbed for Th4+ as a function of adsorption–desorption cycle. [Th4+] 0 =1.71×10−4 mol/L, m/V=1.4 g/L, 298.15 K, pH=2.80±0.05, 5 h References [1]Baybaş D, Ulusoy U (2011) The use of polyacrylamide-aluminosilicate composites for thorium adsorption. Appl Clay Sci 51: 138–146. [2]Avedesian MM, Baker H (1999) ASM specialty handbook Mg and Mg alloys. ASM International, Materials Park, Ohio. [3]Van Horn JD, Huang H (2006) Uranium(VI) bio-coordination chemistry from biochemical, solution and protein structural data. Coord Chem Rev 250: 765–775. [4]Hanif MA, Nadeem R, Bhatti HN, Ahmad NR, Ansari TM (2007) Ni(II) biosorption by Cassia 83
  • fistula (Golden Shower) biomass. J Hazard Mater B13: 345–355. [5]Satapathy D, Natarajan GS (2006) Potassium bromate modification of granular activated carbon and its effect on nickel adsorption. Adsorption 12: 147–154. [6]Wang X, Xia S, Chen L, Zhao J, Chovelon J, Nicole J (2006) Biosorption of cadmium(II) and lead(II) ions from aqueous solutions onto dried activated sludge. J Environ Sci 18: 840–844. [7]Atia AA (2005) Studies on the interaction of mercury(II) and uranyl(II) with modified chitosan resins. Hydrometallurgy 80: 13–22. [8]Humelnicua D, Dinu MV, Dr˘agan ES (2011) Adsorption characteristics of UO 2 2+ and Th4+ ions from simulated radioactive solutions onto chitosan/clinoptilolite sorbents. J Hazard Mater 185: 447–455 [9]Guerra DL, Viana RR, Airoldi C (2009) Adsorption of Th(IV) on chemically modified Amazon clays. J Braz Chem Soc 20: 1164–1174. [10]Guerra DL, Viana RR, Airoldi C (2009) Adsorption of thorium cation on modified clays MTTZ derivative. J Hazard Mater B 168:1504–1511. [11]Unuabonah EI, Adebowal KO, Olu-owolabi BI, Yang LZ, Kong LX (2008) Adsorption of Pb (Ⅱ) and Cd (Ⅱ) from aqueous solutions onto sodium tetraborate-modified Kaolinite clay: Equilibrium and thermodynamic studies. Hydrometallurgy 93: 1–9. [12]Wang C C, Chang CY, Chen CY (2001) Study on Metal Ion Adsorption of Bifunctional Chelating/Ion-Exchange Resins. Macromol Chem Phys 202: 882–890. [13]Biswas M, Mukterjee A (1994) Synthesis and Evaluation of Metal-Containing Polymers. Adv Polym Sci 115: 89–123. 84
  • Recovery of 233 U from irradiated thorium oxide using 5% TBP as extractant Zheng Li, Chunxia Liu, Mumei Chen, Haogui Zhao, Shuhua He, Qingnuan Li, Borong Bao, Yuanxian Xia, Lan Zhang* Shanghai Institute of Applied Physics, Chinese Academy of Sciences * Corresponding author: zhanglan@sinap.ac.cn As the rapid increase of energy demand all over the world, a considerable interest is generated in the use of thorium as a potential source of fissile material in nuclear energy. Compared with uranium-based fuels, the thorium-based reactor fuels have some advantages: 1) thorium is a potentially abundant source of energy, 2) produces less long-lived minor actinides, 3) the control of proliferation of nuclear fuels for use as weapons. To make full use of thorium source, spent fuel reprocessing is a critical step in a fully integrated thorium fuel cycle. An extraction flowsheet using 5% TBP as extractant is studied in our laboratory for recovery 233U from irradiated thorium oxide. Firstly, ThO 2 could be dissolved in the solution of HNO 3 in the presence of F- and Al3+. A batch-dissolution yielding a 430g/L of Th(NO 3 ) 4 solution was achieved in our work. Then, a feed adjustment of concentration of Th(NO 3 ) 4 and HNO 3 was required after dissolution to provide suitable feed solutions for solvent extraction, in which 233U could be separated from 232Th and fission products by extraction in 5% TBP followed by scrubbing and stripping using mixer settler. The recovery of 233U is more than 99.9% and the content of 232Th in 233U is less than 0.04% (separation factor of Th in U is more than 2.5×105). Decontamination factors for main fission products, such as Zr, Nb, Ru, Rh, Ce and Eu, are more than 1.0×105. The recovery of thorium was not considered in the 5% TBP extraction flowsheet, if necessary, the thorium-based spent fuel should be treated with Thorex process using 30% TBP as extractant. Additionally, the concentration and purity of 233U product obtained from solvent extraction could be further improved by an ion-exchange process. 1AF 1AX 5%TBP/dodecane 1AS Th:50g/L U:0.5g/L HNO3 1AU 7 Trace Th in U mass 5 1CX HNO3 1A 1AW 6 Trace U in Th mass 1C 1CU 1CW Trace Th in U mass Trace U Fig.1. extraction flowsheet of recovey of 233U from irradiated thorium oxide using 5% TBP as extractant 85
  • Synthesis and Characterization of UO 2 2+-ion Imprinted Polymer for Separation and Preconcentration of Trace Uranyl ions Meng Hua,b, Li Zhenga, Jia Linaa,b, Jiang Daweia,b, Zhang Lana,*, Zhou Weia a Shanghai Institute of Applied Physics, Chinese Academy of Sciences b University of Chinese Academy of Sciences, Chinese Academy of Sciences * Corresponding author: zhanglan@sinap.ac.cn Molecular imprinting technology (MIT) has been studied extensively due to its high selectivity and good affinity for the target molecule. As a branch of molecular imprinted polymer, ion-imprinted polymer(IIP) is a promise material which are capable of ions recognition. Various metal ions such as UO 2 2+[1-4], Th4+[5], Cu2+[6], Cd2+[7] has been used as the template ion of IIPs, among of which bulk polymerization is the most commonly used imprinting technique.However, IIPs prepared by this method have been encountered with various problems such as nonuniform particle size and shape, inaccessible binding sites situated deep inside the bulk of the polymer matrix. So, surface imprinting technology where the IIPs layers are coated onto substrates was developed. We herein report a novel uranyl ion surface imprinted polymer material prepared by forming ternary complexes of uranyl ion with acrylamide and maleic anhydride functionalized silica particles followed by surface imprinting with 2-hydroxy ethyl methacrylate, ethylene glycol dimethacrylate, AIBN and 2-methoxyethanol as the functional monomer, cross linker, initiator and porogen respectively. HCl solution (3 mol/L) was used to leach out the uranyl ion from the IIP particles.Control polymer(CP) was also prepared under the identical conditions without the uranyl ion. Both IIP and CP particles were characterized by Fourier Transformed Infrared Spectroscopy(FT-IR) and scanning electron microscopy (SEM). FT-IR spectra of unleached and leached IIP material was showed in Fig.1. The vibration frequency due to UO 2 2+ at ~923 cm-1 have been observed in unleached IIP particles [8]. The absence of the vibration frequency at ~923 cm-1 in the leached IIP proves the removal of uranyl ion while leaching. SEM image of ion-imprinted polymer was showed in Fig.2 in which a porous and rough structure was observed on the surface of imprinted polymer as a result of surface imprinting reaction. Higher adsorption capacity of the imprinted polymer was observed than that of control polymer in the batch extraction mode. The maximum adsorption capacity of the imprinted polymer was calculated as 14.65mg/g for 15mg/L of uranium at pH 6.0. Selectivity of the prepared IIP toward uranyl ion in the presence of other selected metal ions is to be investigated [1]M. Shamsipur, J. Fasihi and K. Ashtari, Anal. Chem. 79, 7116 (2007). [2]T. E. Milja, K. P. Prathish and T. P. Rao, J. of Hazard. Mater. 188, 384 (2011). [3]V. E. Pakade, E. M. Cukrowska, J. Darkwa, G. Darko, N. Torto and L. Chimuka, Water Science & Technology 65, 728 (2012). [4]S. Sadeghi and E. Aboobakr, Microchim. Acta 178, 89 (2012). [5]X. Z. Ji, H. J. Liu, L. L. Wang, Y. K. Sun and Y. W. Wu, J. Radioanal. Nucl. Chem. 295, 265 (2013). [6]S. C. L. Pinheiro, A. B. Descalzo, I. M. Raimundo Jr., G. Orellana and M. C. Moreno-Bondi, Anal. Bioanal. Chem. 402, 3253 (2012). [7]G. Z. Fang, J. Tan and X. P. Yan, Anal. Chem. 77, 1734 (2005) [8]C. R. Preetha, J. M. Gladis and T. P. Rao, Environ. Sci. Technol. 40, 3070 (2006). 86
  • Fig. 1. FT-IR spectra of (a) unleached IIP and (b) leached IIP3. Fig. 2. SEM image of ion-imprinted polymer 87
  • Solid phase extraction using N-doped carbonaceous covalent organic frameworks for treatment of uranium (VI) ions from water solutions Chiyao Bai, Xiaoyu Yang, Songbai Liu, Juan Li, Shuang Zhang, Ying Huang, Kecheng Cao, Xiaosheng Zhao, Lijian Ma*, Shoujian Li* College of Chemistry, Sichuan University, Chengdu 610064, PR China. * Corresponding author: sjli000616@scu.edu.cn ma.lj@163.com Uranium is a well known key nuclide in nuclear fuel cycling. Over the past few decades, lots of work has been done to selective separation of uranium present in nuclear fuel effluents, mine tailings, seawater, and other waste sources [1]. In this study, N-doped carbonaceous covalent organic frameworks (CCOF) which is effective for the extraction of UO2+ in the 2 simulated nuclear fuel effluents. To prepare the CCOF, the pre-synthesised N-containing covalent organic framework act as simultaneous C and N source to directly prepare CCOF by “segregated” microwave irradiation, and the acid/base stability and the γ-irradiation stability was checked using HCl and NaOH solution with different normalities (6, 9 N) for 7 day and irradiated at ambient temperature by 60Co gamma source (dose ratio: 6.2 kGy/h) in the dose-range of 10-100 kGy. The characterization of CCOF was examined by SEM (Scanning Electron Microscope), FI-IR (Fourier Transform InfraRed spectrometer), N2 adsorption and elemental analysis of CHN. The elements analysis shows that the nitrogen contents of CCOF is 12.9 wt%, and the bonds at 2222 cm-1 in the FT-IR spectrum can be assigned to the stretching vibration of a C≡N bond [2], which is the unique aspects of the as-prepared CCOF. The nitrogen adsorption/desorption isotherm show that CCOF is a type of substantially microporous materials [3], and the BET surface of CCOF is 507 m2 g-1. The pore size distribution plots obtained by the Horvath-Kawazoe (HK) mode showed that carbon product has narrow pore size distribution, with the peak maxima at diameters of 0.4 nm. Furthermore, all of the characteristic FI-TR peaks of the starting materials remained the same after γ-irradiation and acid/base treatment and no extra peaks were observed, exhibiting these materials’ outstanding radiation stability and chemical stability toward acid and base. The batch sorption experiments were performed to examine the separation performance of CCOFs for uranium(VI) in a simulated nuclear effluent containing 12 co-existent cations with different pH value (4.5, 2.5 and 1). The results in Fig.4 demonstrated the high selectivity of CCOF-SCU1 for uranium(VI). More interestingly, the amount of uranium adsorbed accounts for unreported 80% of the total adsorption amount at pH 1with uranium(VI) sorption capacity of 50 mg g-1. This type of ultramicroporous carbon materials not only has great potential for the separation and recovery of uranium from uranium-containing aqueous solution, but even could be applicable to various fields, such as heavy metal removal. [1] T. P. Rao, P. Metilda,and J. M. Gladis, Talanta 68, 1047 (2006). [2] C. R. Pérez, S. H. Yeon, J. Ségalini, V. Presser, P. L. Taberna, P. Simon, and Y. Gogotsi, Adv. Funct. Mater. 23, 1081 (2013). [3]Y. S. Yun, S. Y. Cho, J. Y. Shim, B. H. Kim, S. J. Chang, S. J. Baek, Y. S. Huh, Y. S. Tak, Y. W. Park, S. J. Park, and H. J. Jin, Adv. Mater. 25, 1993 (2013). 88
  • Fig. 1. FIIR spectra of CCOF-SCU1 treatment with 9 N HNO3: (a) before treatment, (b) treatment for 1 day, and (c) treatment for 7 days. Fig. 2. FIIR spectra of CCOF-SCU2 treatment with 9 N HNO3: (a) before treatment, (b) treatment for 1 day, and (c) treatment for 7 days. Fig. 3. The molar ratio of uranium to the Fig. 4. The molar ratio of uranium to the sum of all metal ions, and (inset) the sum of all metal ions, and (inset) the competitive adsorption capacity of competitive adsorption capacity of coexistent on CCOF-SCU2. coexistent on CCOF-SCU1. 89
  • Extraction of Thorium(IV), Uranium(VI) and Rare Earths with NTAamide Dong-Ping Sua, Huang Huanga,b, Song-Dong Dinga,*, Shou-Jian Lia, Ning Liub a College of Chemistry , Sichuan University , Chengdu 610064 , China b Key Laboratory of Radiation Physics and Technology, Ministry of Education, Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610064, China * Corresponding author: dsd68@163.com Thorium is three times more abundant as compared to uranium in nature and is an easily exploitable nuclear resource. In recent years, molten salt reactor (MSR) as one of the Generation-IV nuclear power has been proposed for thorium utilization. With the development of TMSR, thorium fuel cycle becomes an important element for the long-term nuclear power growth. For reprocessing of spent Th-based fuels, THOREX process, based on solvent extraction of uranium and thorium with tributyl phosphate (TBP), has so far been the most viable route. However, for TBP, due to its mediocre irradiation stability, relatively high solubility in water, and easy occurrence of the third phase in the case of the extraction of tetravalent actinides, there are still many problems to be solved for THOREX process. Therefore, the exploration of the alternate for TBP is very essential. Compared with TBP, the amide extractants have some significant advantages. The primary radiolytic and hydrolytic fragments are amines and carboxylic acids, which are tolerable without serious interference in the extraction process. In addition, the amide compounds consist only of C, H, O and N atoms and can be incinerated to gaseous products, generating lesser secondary waste. In the present paper, a derivative from NTA (NTAamide, Fig. 1) was synthesized and its extraction behaviors have been investigated by using kerosene as diluent. The effect of HNO 3 concentration on the extraction was shown in Fig. 2. The value of D Th decreases with the increase of acidity until it levels on 0.5 mol/L and subsequently increases with a further acidity increase. NTAamide displays a great selectivity for Th over U and REs, suggesting a good prospect for the separation of Th and U. Slope analysis (Fig. 3) indicates metal ions are extracted as mono-solvated species by NTAamide. Fig. 1. The structure of NTAamide. Fig. 2. Effect of HNO3 concentration on distribution ratio of Thorium(IV), Uranium(VI) and Rare Earths. 90 Fig. 3. Effect of NTAamide concentration on distribution ratio of Thorium(IV), Uranium(VI) and Rare Earths.
  • Adsorption and Separation Characteristics of Thorium from Nitric Acid Solution Using Silica-Based Anion Exchange Resins Yanliang Chena, Long Zhao a *, Linfeng Heb, Yuezhou Weia, Fangdong Tangb a Nuclear Chemical Engineering Laboratory, School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, P. R. China b Chemical and Ionizing Radiation Metrology Institute, Shanghai Institute of Measurement and Testing Technology, Shanghai 201203, P.R. China Corresponding author: Tel/Fax +86-21-34207654; E-mail: ryuuchou@sjtu.edu.cn. Today's reactors burn essentially uranium-235 and they only use about 1% of the natural uranium [1]. For this reason, uranium reserves cannot supply the needs of reactor operation for quite a long time. Thorium is 3~4 times more abundant worldwide than uranium. What's more, the thorium-uranium nuclear fuel cycle, in which the main fissile nucleus is uranium-233 and fuel regeneration is ensured through neutron capture on thorium-232 offers many potential advantages [1-2]. The usage of thorium in future reactors is quite feasible. On the other hand, in traditional nuclear fuel cycle, the application of advanced silica-based anion exchange resins has been studied widely. Compared to the commercial polymer-based anion exchange resins, the silica-based anion exchangers are characterized by a rapid ion exchange rate, relatively excellent radiation-resistance and significantly low pressure loss in a packed column [3]. This work mainly discusses the adsorption behavior of thorium using silica-based anion exchange resins and the separation of thorium and uranium with column separation methods. Table I. Characterization of the silica-based anion exchange resins Pore size of Adsorbent silica (nm) Size Functional group (μm) SiPyR-N3 50 37~74 SiPyR-N4 600 37~74 AR-01 600 37~74 In order to separate thorium form nitrate acid solutions, several silica-based anion exchange resigns were prepared, including weak base anion exchange resin named SiPyR-N3, strong base anion exchange resin named SiPyR-N4 and an kind of anion exchange resin with complex functional groups named AR-01. Their structures and some information are shown in Table 1. Batch experiments were carried out to investigate the adsorption behavior of thorium from nitric acid solutions. The influence of the concentration of nitric acid and Th, contact time, and the effect of coexist metals and nitrate sucha as NaNO3 and Mg(NO3)2 was investigated in detail. As shown in Fig.1, at high concentration of nitric acid, the resins exhibit higher adsorption ability for thorium. In addition, it is found that better selection for thorium from accompanying metal ions is achieved in higher nitric acid concentration. The adsorption kinetics was well described by the pseudo-second order model equation, and the 91
  • adsorption isotherm was better fitted by the Langmuir mode. The maximum Th uptake of SiPyR-N4 was calculated as 27.3mg/g-resin. Lastly, nitrates, such as NaNO3 and Mg(NO3)2, were found to promote the adsorption of thorium obviously. 40 q / mg/g 30 AR-01 N4 N3 20 10 0 0 3 6 [HNO3] / M 9 Fig.1.Adsorption of thorium in different concentrations of nitric acid In the thorium/uranium fuel cycle, the separation of thorium and uranium is the key step. According to the batch adsorption result, SiPyR-N4 was selected to use in the column experiments. The results of batch experiment show that the separation factor between thorium (IV) and uranium (VI) is about 10 at 9M HNO3. Therefore, the column separation experiment was carried out at this concentration of nitric acid. It was found that thorium comes to a breakthrough of the column at the point BV=11 when the SV=12 (see Fig.2). Fig.3 exhibits the column chromatographic experiment result. As seen, thorium and uranium were separated effectively. 40 10 Concentration / mM Leakage concentration / mM Dead volume 8 6 4 2 0 0 5 10 15 Bed volume / ml/ml 30 Fig.2.Breakthrough curve of thorium 9M HNO3 0.1M HNO3 H2O Th U 20 10 0 0 20 Feed 50 100 150 Effluent volume / ml 200 Fig.3.Column chromatographic experiment [1] S David, E Huffer, H Nifenecker. Europhysics News, 2007. [2] k Anantharaman, V Shivakumar, D Saha. Journal of Nuclear Materials. 383 (2008), 119~121. [3] Y Wei, M Kumagai, Y Takashima, M Asou, T Namba, K Suzuki, A Maekawa, S Ohe. Journal of Nuclear Science and Technology. Volume 35, Issue 5, 1998, 357~364. 92
  • Adsorption and elution of rhenium (VII)with a porous silica-based anion exchanger AR-01 Xiaolong Wang, Jianhua Zu, Yuezhou Wei* School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, China * Corresponding author: yzwei@sjtu.edu.cn. Rhenium is an extremely rare metal with high melting point. It plays an important role in chemical and metallurgical industries. However, the abundance of rhenium is quitelow and there is no independent rhenium mineral in nature. In recent years, effective recovery of rhenium hasbecome an interesting field of study. Currently there are various methods to recover Re(VII) including extraction, precipitation and ion exchange. Ion exchange method is relatively a simple, convenient and environment-friendly method [1]. AR-01 resin is a novel silica-based anion exchanger with both weak and strong basic groups. In this paper, the adsorption and elution behavior of of Re(VII) by AR-01 resin is evaluated in detail. Compared with traditional anion exchange resins, the main advantage of AR-01 resin is its high resistance to irradiation environment [2], thus it is potential to be applied for nuclear waste processing. The results will also offer important information for the applicability of AR-01 resin for the separation of technetium, because of their similar chemical prosperities. The influence of acid concentration, the adsorption kinetic and the adsorption capacity were studied by the batch experiment. The static saturated adsorption capacity is 333mg g-1 resin at pH 2.0 in nitric acid at 298K with equilibrium time less than 10 min. Fig. 1 shows the breakthrough curves,the dynamic saturated adsorption capacity is 312 mg g-1 resin. In condition of pH 2.0 and rhenium concentration 2.0 mM, all the rheniumpassed the resinwas absorbed before the breakthrough point. Fig. 2 shows the elution curve at flow rate 0.2 ml min-1using 8%NH 3 ·H 2 O+10%NH 4 NO 3 . 12 bed volumes of elute solution can elute 90% of rhenium and 30 bed volumes of elute solution is required to elute 99.3 % of rhenium absorbed by the resin. The resin can be regenerated and reused without decrease in adsorption capacity.AR-01 resin is proved to be potential for the recovery of rhenium with advantages such as large adsorption capacity, higher adsorption and desorption rates. [1] Z.-N. Shu and M.-H. Yang, Chim. J. Chem. Eng., 18 (2010) 372. [2] Y.-Z. Wei, M. Kumagai, Y. Takashima, et al., J. Nucl. Sci. Technol., 35(1998)357. 1.2 1.0 0.2ml/min 1.0ml/min 12000 C/C0 [Re] (ppm) 0.8 0.6 0.4 6000 3000 0.2 0.0 0 9000 50 100 150 Bed volumes 0 0 200 10 20 30 Bed volumes 40 50 Fig. 2.Rhenium elution from AR-01 resin using 8%NH3·H2O+10%NH4NO3. Flow rate 0.2 ml min-1. Fig. 1.Breakthroughcurves and column saturation for Re adsorption at various flow rate. pH 2.0, Re(VII) concentration 2 mM 93
  • Study on the Properties of isoBu-BTP/SiO2-P Adsorbent in the Separation of Minor Actinides Xinpeng Wanga, Ruiqin Liua,Shunyan Ninga, Yuezhou Weia*, Youqian Dingb, Jinling Yangb, Weiqun Shic, Zhifang Chaic a Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai, 200240, China b China Institute of Atomic Energy, Beijing, 102413, China c Institute of High Energy Physics, Chinese Academy of Science, Beijing 100049, China * Corresponding author: yzwei@sjtu.edu.cn The selective separation of the minor actinides (neptunium, americium and curium) and their subsequent transmutation into shorter-lived or stable nuclides in advanced nuclear reactors could have a significant beneficial impact on minimizing the long-term radiotoxicity of HLLW in a final geological repository. However, the actinide(III)-lanthanide(III) intergroup separation from HLLW is not easily achievable owing to their chemical similarities and comparable ionic radii. N-donor molecules of the BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)pyridine) type are widely investigated due to their high selectivity for minor actinides and favorable properties. In this study, we prepared novel isoBu-BTP/SiO2-P adsorbent by impregnating isoBu-BTP extractants into a macroreticular styrene-divinylbenzene copolymer which is immobilized in porous silica particles with a mean diameter of 50μm. The adsorption performance of 241Am(III), lanthanides and some typical fission products was investigated. The influence of nitric acid concentration, the adsorption kinetics and the adsorption capacity were also studied by the batch experiment. It was found that 241Am(III) could be effectively separated from 152Eu(III) by using isoBu-BTP/SiO2-P adsorbent in nitric acid solution. With the increase of initial NO3- concentration the separation factor between 241 Am/152Eu increased dramatically. [1] Y.-Z. Wei, Kanwal Nain Sabharwal, Mikio Kumagai, Toshihide Asakura, Gunzo Uchiyama and Sachio Fujine, Journal of Nuclear Science and Technology, Vol. 37, No.12, p. 1108-1110. [2] Y.-Z. Wei, H. Hoshi, M. Kumagai, T. Asakura, Y. Morita, Journal of Alloys and Compounds 374 (2004) 447-450. [3] H. Hoshi, Y.-Z. Wei, M. Kumagai, T. Asakura, and Y. Morita, Journal of Alloys and Compounds , 408–412 (2006), 1274–1277. [4] WEI YueZhou, WANG XinPeng, LIU RuiQin, WU Yan, USUDA Shigekazu & ARAI Tsuyoshi, Sci China Chem, September 2012 Vol.55 No.9: 1726–1731. 1.8x104 5x103 Am-241 Eu-152 1.2x104 Kd (cm3/g) Kd (cm3/g) 4x103 Am-241 Eu-152 1.5x104 3x103 9.0x103 2x103 6.0x103 1x103 3.0x103 0 0.0 0 1 2 3 HNO3 concentration (mol/L) 0 4 Fig. 1. Element distribution coefficients of 241 Am(III) and 152Eu(III) as a function of the initial HNO3 concentration (298 K, phase ratio: 0.1g/5cm3, tracer (241Am(III), 152 Eu(III)), contact time: 24 h). 94 1 2 3 NO3- concentration (mol/L) 4 Fig. 2. Element distribution coefficients of 241 Am(III) and 152Eu(III) as a function of the initial NO3- concentration (298 K, phase ratio: 0.1g/5cm3, tracer (241Am(III), 152 Eu(III)), contact time: 24 h).
  • Removal of Th4+ Ions from Aqueous Solutions by Graphene Oxide Ning Pana, Yongdong Jina, Chuanqin Xiaa,b,* a College of Chemistry, Sichuan University, Chengdu, 610064, P. R. China b MOE Key Laboratory of Radiation Physics and Technology, Institute of Nuclear Science and Technology, Sichuan University, Chengdu, 610064, P. R. China. * Corresponding author: xiachqin@163.com. Graphene oxide (GO) [1] is prepared through the exfoliation of the oxidized graphite, and can incorporate various oxygen-containing functional groups (hydroxyl, epoxide, carbonyl, carboxyl groups) on its surface. The presence of these oxygen-containing functional groups is well-suited for interaction with cations, which also makes GO hydrophilic and highly negatively charged in water [2]. Additionally, GO has a large area of contact between its sheet-like nanostructure and metal ions. In this report, the removal of Th4+ ions from aqueous solutions was investigated using GO as a adsorbent which was prepared by the modified Hummers' method through batch adsorption experiments at room temperature. Structural characterizations of the adsorbent were also investigated. The influences of the pH value of solution, contact time, adsorbent dose, ionic strength, the initial metal ion concentration and temperature on the adsorption of Th4+ were also investigated. These results indicated that the adsorption of Th4+ was dependent on the pH and independent on the ionic strength and the adsorbent provided significant Th4+ removal (>98.7 %) at pH 3.0 and the adsorption equilibrium was achieved after only 10 min. The Langmuir adsorption isotherm fit the absorption profile very closely, and indicated that a maximum adsorption capacity of 1.77 mmol/g of GO (411 mg/g) after 2 h. The thermodynamic parameters showed that this adsorption process was endothermic and spontaneous. Moreover, the desorption level of Th4+ from GO, by using 0.1 mol/L H 2 SO 4 as a stripping agent, was 84.2 ± 1.2 %, and that of 0.5 mol/L HNO 3 as a stripping agent, was 79.8 ± 3.0 %. [1] W. S. Hummers, R. E. Offeman, Preparation of graphite oxide, J. Am. Chem. Soc., 80 (1958):1339 [2] D Li, MB Müller, S Gilje, RB Kaner, GG Wallace, Processable aqueous dispersions of graphene nanosheets, Nat. Nanotech., 3 (2008): 101-105 Fig. 2. Comparison of experimental and calculated data for two models for the adsorption of Th4+ ions on GO. The calculated data were obtained by applying the Langmuir and Freundlich isotherm equations. Fig. 1. Typical tapping-mode AFM image (top) of GO on mica and the relative height profile (bottom)of the sample. 95
  • Influence of γ-irradiation on the isoBu-BTP/[C2mim][NTf2] extracting system during Dy(III) extraction Weijin Yuana, Yinyong Aob, Long Zhaoa*, Yuezhou Weia, Maolin Zhaib a Nuclear Chemical Engineering Laboratory, School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, P. R. China b Beijing National Laboratory for Molecular Sciences (BNLMS), Department of Applied Chemistry, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, China Corresponding author: Tel/Fax +86-21-34207654; E-mail: ryuuchou@sjtu.edu.cn. Lipophilic Bistriazinylpyridines (BTPs), as heterocyclic nitrogen donor ligands, present attractive selectivity for Am (Ⅲ) over Ln (Ⅲ) that could make them good candidates for minor actinides separation[1]. While, the room-temperature ionic liquids (RTILs) are considered as promising solvents for liquid-liquid extraction due to their physical and chemical properties, such as negligible vapor pressure, good dissolving ability [2]. Thus, BTPs/RTILs extracting system is expected to enhance the efficiency for the separation process. Due to the high radioactive environment of spent fuel reprocessing, the radiation stability of the novel extracting system should be investigated before their practical application. In this work, the radiation effect on the BTPs/RTILs novel extracting system during extraction was investigated. The novel extracting system is made up of 2,6-bis(5,6-dibutyl-1,2,4-triazin-3-yl) pyridine (isoBu-BTP), which acted as extractant, can separate lanthanides and minor actinides effectively in nitric acid solution, and 1-ethyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([C2 mim][NTf2]),which acted as extraction medium, shows a great solubility against isoBu-BTP. It has been reported that Dy(III) possessed similar adsorptive behavior to the MA, and never appeared in spent fuel [3]. For this reason, Dy(Ⅲ), which acted as extract, was chose as an alternative of MA (Ⅲ) in current research H F CH2CH3 N N H3C N N F F N O C C F N N N H N H F O F S S N O O Scheme 1 The structure of isoButhyl-BTP (left) and [C2mim][NTf2](right) In order to elucidate the radiation effect on the extraction, we designed three kinds of extraction: (a) using irradiated [C2mim][NTf2] in combination with unirradiated isoButhyl-BTP; (b) using unirradiated [C2mim][NTf2] in combination with irradiated isoButhyl-BTP; (c) using irradiated isoBuBTP/[C2mim][NTf2]. It was found that extraction ability has been maintained by using unirradiated [C2mim][NTf2] in combination with irradiated isoButhyl-BTP at dose of 500kGy; While an unexpected great increase of Dy(III) extraction was found in the presence of irradiated [C2mim][NTf2] in the same dose (see Fig.1). Irradiated Ionic liquid seems to be the dominant and favorable factor which affected the extraction process. However further experiment revealed that the increase of extraction under irradiation was a false impression. The removal of Dy(III) in current case was not by extraction but by precipitation. Various radiolysis species such as HF, H2SO3, CF3SOONH2 and CF3SOOH generated from irradiated [C2mim][NTf2] were confirmed (see Fig.2). Among them, HF was identified as the major radiolytic product by ion chromatography (see Fig.3). The precipitation collected from the interface of 96
  • aqueous phase and ion liquid phase was identified as DyF3 according XPS analysis. (see Fig.4). The combination of Dy3+ and F- leads to the formation of DyF3 precipitation which promote the removal of Dy(III) in aqueous phase. Accordingly, washing irradiated [C2mim][NTf2] with water can easily erase the inconvenience to extraction, and give a simply way of ionic liquid recycling. The principle intention at the outset of the investigation reported here was to assess the feasibility of novel BTPs/RTILs extracting system for minor actinides separation from spent fuel. The radiolysis behavior of BTPs/RTILs extracting system is considerably complicated, especially, when the novel extracting system is exposed to high radioactive waste fluid. The preliminary results obtained in this work merely elucidate the influence of primary water-soluble radiolytic products such as HF on the Dy(III) extraction. Therefore further research on this topic is needed to be done. Fig.1. Dy3+ extraction from aqueous solution by irradiated [C2mim][NTf2] in combination with unirradiated isoBu-BTP. Fig.3. Ion chromatography spectra of [C2mim][NTf2] before and after γ-irradiation at 500 kGy. Fig.2.The 19F NMR spectra of [C2mim][NTf2] before (a) and after γ-irradiation at 500 kGy (b). Fig.4. XPS spectrum of the precipitation collected from the interface of the aqueous phase and ion liquid phase. [1]Panak, P.J. and A. Geist, Chemical reviews. 113(2): 1199-1236(2013). [2]Dai, S., Y. Ju, and C. J. Chem. Soc., Dalton Trans. (8): 1201-1202(1999). [3]Usuda, S., et al., Journal of Ion Exchange. 21: 35-40(2010). [4]Yuan, L., et al., Dalton Transactions. (38): 7873-7875( 2009). [5]Ao, Y., et al., Dalton Transactions.42 (2013). 97
  • Ethanolamine-isocyanate modified graphite oxide for selective solid-phase extraction of uranium Yin Tian, Lijian Ma, Bo Li, Juan Li, Xiaodan Yang, Kecheng Cao, Chiyao Bai, Ying Huang, Shoujian Li* College of Chemistry, Sichuan University, Chengdu 610064, PR China * Corresponding author: sjli000616@163.com Graphite Oxide (GO) was prepared from purified natural graphite by the Hummers method [1]. The new solid-phase extraction (SPE) adsorbent (ETA-iGO) was prepared by using tolylene-2,4-diisocyanate (TDI) and ethanolamine (ETA) to assemble the multidentate ligands in situ on the GO lamellar matrix through a two-step “grafting to” procedure [2,3]. The newly formed multidentate ligand bears both urea- and carbamate-like structures simultaneously. The in situ assembly mechanism of the multidentate ligand was also explored based on characterization studies. Adsorption capability of ETA-iGO toward uranium was investigated systematically including effects of pH, contact time, uranium load, temperature, competitive ions and so on [4-8]. The results from batch adsorption experiments indicate that the maximum adsorption capacity of ETA-iGO was found to be 0.46 mmol/g-1 (298.15K, pH 4.5) in the pure uranyl nitrate solution. In order to investigate the sorption selectivity of the as-synthesized ETA-iGO for uranium (VI), batch adsorption experiments were performed in a simulated nuclear industrial effluent composed of totally 12 main sensible nuclides including uranyl ion, using GO matrix for comparison. The U(VI) adsorption capacity increased significantly from 0.25 mmol·g-1 for GO to 0.35mmol·g-1 for ETA-iGO, and the uranium selectivity of the sorbents used also increased exhilaratingly from 35.4% for GO to 59.6% for ETA-iGO. 98
  • Fig. 1.Competitive adsorption capacity of coexistent ions on GO and ETA-iGO (c 0 = 0.42 mmol·L-1 for all cations, pH = 4.5, t = 120 min, V = 20 mL,T = 293.15 K, and w = 10 mg) [1] W. Hummers and R. Offeman, J. Am. Chem. Soc. 80, 1339 (1958) [2] S. Stankovich, R. D. Piner, S. T. Nguyen and R. S. Ruoff, Carbon 44, 3342–7 (2006) [3] D. Zhang, S. Zua and B. Han, Carbon 47, 2993 (2009) [4] Y. Zhao, C. Liu, M. Feng, Z. Chen, S. Li, G. Tian, L. Wang, J. Huang and S. Li, J. Hazard. Mater. 176, 119 (2010) [5] H. Wang, L. Ma, K. Cao, J. Geng, J. Liu, Q. Song, X. Yang and S. Li, J. Hazard. Mater. 229, 321 (2012) [6] J. Geng, L. Ma, H. Wang, J. Liu, C. Bai, Q. Song, J. Li, M. Hou, and S. Li, J. Nanosci. Nanotechnol. 12, 7354 (2012) [7] Q. Song, L. Ma, J. Liu, C. Bai, J. Geng, H. Wang, B. Li, L. Wang and S. Li, J. Colloid Interface Sci. 279, 307 (2004) [8] J. Liu, J. Li, X. Yang, Q. Song, C. Bai, Y. Shi, L. Zhang,C. Liu, S. Li and L. Ma, Mater. Lett. 97, 177(2013) 99
  • Separation Behavior of Rare Metals by Functional Xerogels Impregnated with MIDOA Extractant a Rana Syed Masuda, Hitoshi Mimuraa,*, Yuji Sasakib, Masaki Ozawac Dept. of Quantum Science and Energy Engineering, Tohoku University, Japan b Japan Atomic Energy Agency, Japan c Tokyo Institute of Technology, Japan * Corresponding author: hitoshi.mimura@qse.tohoku.ac.jp The novel organic extractant, MIDOA (2,2’-(methylimino)bis(N,N-dioctylacetamide)), with high extractability for rare metal ions was microencapsulated by the sol-gel method using alginate polymer matrices. The uptake behaviors of rare metals (Pd(II), Os(IV), Ir(III), Pt(IV), Au(III) and Hg(I)) in the presence of HCl were examined by the batch method. The uptake rate of Os(IV), Pt(IV), Au(III) and Hg(I) was very high. The uptake equilibrium was attained within 1 h with large (~100%) uptake (%). The uptake rate of Pd(II) was fairly fast in the initial stage, 90% uptake within 30 minutes, and equilibrium was attained within 12 hour (Fig. 1). Relatively large Kd values above 104 cm3/g were obtained for Au(III), Os(IV) and Hg(I) ions in a wide range of HCl concentration up to 6 M. The adsorption of rare metal ions followed a Langmuir-type adsorption equation, and the maximum capacity of Pd(II) was estimated to be 0.5 mmol/g. The adsorbed Pd(II) ions were quantitatively eluted with 0.5M thiourea(TU)-1M HCl, indicates strong complex formation ability with TU. Stepwise chromatographic separations of rare metals were investigated using different concentration of TU in 1 M HCl (Fig. 2). The novel extractant, MIDOA, has N atoms in the middle of frame work and amide groups at both edges, and the binding with metal ions is thus strongly affected by the N atoms [1, 2]. The N donor is a soft atom for the coordination in the HSAB principle and has strong binding ability with soft metal ions. So far MIDOA has been well known to have excellent reactivity with oxoanions such as Tc(Re) and Pd, etc.; in the previous papers, the separation behavior of Tc(Re) from HLLW has been clarified by using MIDOA microcapsules in HNO3 system [3]. [1] Y. Sasaki, M. Ozawa, T. Kimura, K. Ohashi, Solvent Extraction and Ion Exch., 27, 378-394 (2009) [2] Y. Sasaki, Y. Kitatsuji, T. Kimura: Chemistry Letters, 36, 11, 1394-1395 (2007) [3] Rana S. Masud, H. Mimura, T. Sugimori, Y. Niibori, Y. Sasaki, Proc. of EMR 2012, Malaga, Spain. 50 100 H2O 60 0.1M TU -1M HCl 0.2M TU -1M HCl 0.5M TU -1M HCl 30 Pd 0 0 2 4 6 8 10 Time (h) 12 14 Pt Au 20 Os Ir 40 Elution (%) Uptake (%) 80 2M HCl 40 Hg 16 18 Pd Os Ir Pt 20 Au Hg 10 0 20 Fig. 1. Uptake rate of Os(IV), Pt(IV), Au(III), Hg(I), Pd(II) and Ir(III) for MIDOA-MC. 0 20 40 60 Volume (cm3) 80 100 120 Fig. 2. Chromatographic elution of Pd(II) from loaded mixed rare metals in MIDOA-MC column. 100
  • Session 6: Pyro-Separation Technologies 101
  • Recent Study on Pyrochemical Treatment of Spent Nitride Fuels in JAEA Hirokazu HAYASHIa,*, Takumi SATOHa, Hiroki SHIBATAa, Masaki KURATAa, Takashi IWAIa, Yasuo ARAIa a Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan * Corresponding author: hayashi.hirokazu55@jaea.go.jp Partitioning and transmutation of long-lived minor actinide (MA) is an option for the future nuclear fuel cycle program to reduce the burden of the high-level radioactive waste and to use the repository efficiently. Nitride fuel cycle for transmutation of MA has been developed in Japan Atomic Energy Agency (JAEA) under the double-strata fuel cycle concept [1,2]. The double-strata nuclear fuel cycle consists of the commercial power reactor fuel cycle and the dedicated transmutation cycle using the accelerator-driven system (ADS) with the MA nitride fuels. Nitrides are suitable for the fuel materials for transmutation of MA because they support hard neutron spectrum, indicate high thermal conductivity and high metal density like metallic fuels, and high melting point and isotropic crystal structure like oxide fuels [3]. Pyrochemical process is proposed for the reprocessing of spent nitride fuel, since it has several advantages over wet process in treating nitride fuel for transmutation, such as compactness of the facility, resistance to radiation damage, margin of criticality, and recycling feasibility of 15N. We have developed elemental technologies of the main processes such as molten salt electrorefining of actinide nitrides and renitridation of recovered actinides [1,2]. Other processes are similar to the technology for the metal fuel treatment and have been studied elsewhere. The dedicated fuel for ADS is U-free fuel and an inert matrix material is added in place of U. We proposed to use ZrN as an inert matrix material as the first candidate. Solid solution type (An,Zr)N fuels can be fabricated with a wide range of composition. The proposed nitride fuel cycle for ADS aims to transmute long-lived MA into short-lived or stable nuclides. Nitride fuels for the first transmutation cycle are fabricated from MA recovered from the high-level liquid waste (HLLW) of the commercial cycle. Neutronic study of the ADS system led to using the nitride fuel consisting of PuN+MA nitride and ZrN as an inert matrix [4]. We have developed basic concept of the main process for pyrochemical reprocessing of spent nitride fuels. For this purpose, knowledge on the behavior of each element in the pyrochemical process is necessary. Our experimental study contains the behavior of pure actinide nitrides, zirconium nitride, lanthanide nitrides, and the solid solutions of actinide nitrides and other components such as ZrN and FPs. In this study, we summarized our activity on developments of the pyrochemical treatment of the spent nitride fuels. [1] T. Mukaiyama, et al., “Review of Research and Development of Accelerator-Driven System in Japan for Transmutation of Long-lived Nuclides,” Prog. Nucl. Energy, 38, 107 (2001). [2] K. Minato, Y. Morita, T. Kimura, H. Oigawa, Y. Arai, T. Sasa, "Recent Research and Development on Partitioning and Transmutation by “Double-strata Fuel Cycle Concept” in JAEA," Proc. GLOBAL 2009, Sep. 6-11, 2009, Paris, France, 504-512 (2009) [CD-ROM]. [3] Y. Arai, “3.02 Nitride Fuel,” in Comprehensive Nuclear Materials, Elsevier (2012). [4] K. Nishihara, K. Iwanaga, K. Tsujimoto, Y. Kurata, H. Oigawa, T. Iwasaki, ‘‘Neutronics Design of Accelerator-driven System for Power Flattening and Beam Current Reduction,’’ J. Nucl. Sci. Technol., 45, 812 (2008). 102
  • Thorium based Molten Salt Fuel Cycle Qing-nuan Li*, Lan Zhang, Wen-xin Li,Guo-zhong Wu Shanghai Institute of applied physics, Chinese Academy of Sciences, Shanghai, China *Corresponding author. Tel: +86 21 39194059. Email: liqingnuan@sinap.ac.cn In 2011, Chinese Academy of Sciences (CAS), after discontinuing the research and development activity in nuclear energy for decades, started to implement Strategic Priority Research Program "Future Advanced Fission Nuclear Energy (FANE)". To perform this program, two sub-bases, the north and the south, were deployed in CAS. Shanghai Institute of applied physics (SINAP), as the south sub-base, is taking in charge of research and development of "Thorium-based Molten Salt Reactor Nuclear Energy System (TMSR)". According to this research plan, two kinds of molten salt nuclear reactors, i.e. 2MW uranium-thorium fluoride molten salt reactor and 2MW pebble bed fluoride salt-cooled high temperature reactor (PB-FHR), will be designed and developed. Three fuel cycle models will also be implemented orderly, one-through fuel cycle on PB-FHR, modified open fuel cycle on PB-FHR and TMSR, and closed fuel cycle on TMSR. Pyrochemical processing methods are judged to be the only technologies for the fuel of MSRs with integrated reprocessing technologies. Because the liquid fuel for MSR is a mixture of molten fluorides, the fuel processing and reprocessing technologies planned are pyrochemical or pyrometallurgical techniques, which are based on separation of 233U and fission products in molten fluoride salt. Considering the special advantages of fluoride volatility and electrometallurgical techniques, a preliminary protocol based on closed fuel cycle has been proposed for the treatment of fuel from TMSR. The recycling techniques of fuels proposed in this protocol include fluoride volatilization, distillation of molten salt carriers, electrochemical deposition. The simple experimental devices for above techniques have been established, and the feasibility studies are ongoing in SINAP. 103
  • The study on the solubility of rare earth oxides in a new molten salts LiCl-NaCl-MgCl 2 Yingcai Wang, Wei Han, Milin Zhang, Mei Li*, Yaochen Liu Key Laboratory of Superlight Materials and Surface Technology, Ministry of Education, College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China * Corresponding author: meili@hrbeu.edu.cn. Molten salts, and particularly molten chlorides, are well known as good reaction media for performing selective solubilization or precipitation in chemical reactions, and have already been proposed as a promising route for the treatment of raw materials [1]. Pyrochemical separation processes in molten media have more recently been proposed as a promising option in the nuclear fuel cycle for the future [2, 3]. The separation of actinides from lanthanides—/the more difficult fission products to separate due to their similar chemical properties—from oxide nuclear fuels, which involves the dissolution of some of the products in a molten chloride [4]. As we know, rare earth oxides (RE 2 O 3 ) are usually insoluble in molten chlorides, such as LiCl-KCl, NaCl-KCl, CaCl2 -NaCl, and so on. In order to improve the solubility of RE 2 O 3 , a new salt system, LiCl-KCl-MgCl 2 melts, was selected as a solvent system. The solubility of RE 2 O 3 in LiCl-KCl-MgCl 2 melts was researched. Selection of molten salt system has numerous selection criteria, such as melting temperature, solubility of metallic species, electro-activity range, salt stability, salt volatility, and so on. It is important to choose salt mixtures as solvent medium in order to reduce the salt volatility and increase the solubility of RE 2 O 3 in the molten salt system. The volatilization loss of solvent salt in different salts system i.e. the LiCl-NaCl-MgCl 2 melts and KCl-LiCl-MgCl 2 melts were determined. The details as follow: The mixture of an anhydrous LiCl-NaCl(KCl)-MgCl 2 (6.0: 3.2: 1.5 mass) powder in a corundum crucible and heated to specified temperature in electric furnace under argon atmosphere. The volatile loss in LiCl-NaCl-MgCl 2 and LiCl-KCl-MgCl 2 systems were calculated by the measuring the weight loss of melts in every 30 min using an electronic balance in the temperature range from 873 to 1073 K. The comparison of volatile loss in these two different systems is shown in Fig. 1. The results show that the volatile loss in LiCl-KCl-MgCl 2 molten salts is much larger than that in LiCl-NaCl-MgCl 2 , which proved that LiCl-NaCl-MgCl 2 melts are more stable than LiCl-KCl-MgCl 2 melts in the experimental temperature range. The volatile was collected by an inverted glass funnel above the crucible containing LiCl-KCl-MgCl 2 salts, and analyzed by XRD (Fig. 2). The result indicates that volatile from LiCl-KCl-MgCl 2 salts is consist of KCl and KMgCl 3 . Based on the experimental results, the salt mixtures LiCl-NaCl-MgCl 2 were selected as a solvent system. After the addition of RE 2 O 3 in LiCl-NaCl-MgCl 2 melts, we took samples from the melt, diluted them in water and then analyzed by ICP-AES. Fig. 3 shows the content of rare earth elements in LiCl-NaCl-MgCl 2 melts in the temperature range from 873 to 1073 K. Due to the insolubility of the RE oxides or oxychlorides in water, it can be confirmed that there is RE (III) in LiCl-KCl-MgCl 2 -RE 2 O 3 melts. The RE (La, Pr, Ho, Yb, Y) oxides have the higher solubility in LiCl-NaCl-MgCl 2 melts, but the solubility of Er 2 O 3 is lower significantly. We believe that there is a reaction between Re 2 O 3 and MgCl 2 . When Re 2 O 3 was added 104
  • into the LiCl-NaCl-MgCl 2 melts, the solid RE 2 O 3 is chlorinated by MgCl 2 , and forms RECl3. The reaction as follow: RE 2 O3 (s)+3MgCl2 (l)=2RECl3 (l)+3MgO(s) (1) Using HSC software, the values of Gibbs energy of the chlorinated reaction were calculated (Fig. 4). In general,the value of Gibbs free energy of the reaction increases with increasing temperature, so the increase of temperature does not help to the reaction. The Gibbs free energy of the chlorinated reaction of CeO 2 become significantly larger, and the values of Gibbs free energy of the chlorinated reaction of heavy rare earth oxides are greater than zero, but the chlorinated reaction can still occur. The reason may be that the product of MgO is insoluble and deposit in the molten salts, which leads to the equilibrium reaction move to the products—generate rare earth chlorides. Figure 1 The weight loss of melts. Figure 2 XRD pattern of volatile salt. Figure 3 The solubility of rare earth oxides. Figure 4 Gibbs free energy of reaction (1). • [1] D.M. Ferry, G.S. Picard, B.L. Tr e milllon, Trans. Instn. Min. Metall., 97 (1988) C21-C30 (Section C: Mineral Process Extr. Metall.). [2] T. Koyama, M. Iizuka, H. Tanaka, M. Tokiwai, Y. Shoji, R. Fujita, T. Kobayashi, J. Nucl. Sci. Technol., 34 (1997) 384. [3] T. Inoue, Y. Sakamura., in: M. Gaune-Escard (Ed.), Molten salts: from fundamentals to applications, NATO science series II, vol. 52, Kluwer Academic Publishers, Dordrecht, 2002, pp. 249-261. [4] Y. Castrillejo, M.R. Bermejo, E. Barrado, A.M. Martínez, P. Díaz Arocas, J. Electroanal. Chem., 545 (2003) 141-157. 105
  • Separation of SmCl3 and DyCl3 by galvanostatic electrolysis in LiCl-KCl melts at magnesium electrodes Yusheng Yang, Milin Zhang, Wei Han*, Bin Liu, Pengyuan Sun Key Laboratory of Superlight Materials and Surface Technology, Ministry of Education, College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China * Corresponding author: weihan@hrbeu.edu.cn. zhangmilin@hrbeu.edu.cn. Rare earth elements, especially samarium, because of the big neutron capture section, are harmful for the nuclear fuel cycle [1, 2]. Due to the similar properties between lanthanides and actinides, lanthanides are difficult to be separated from spent nuclear fuel. Pyrometallurgy reprocessing of spent nuclear fuel is now considered as one of the most promising options for an innovative nuclear fuel cycle [3]. In the pyrochemical separation processes, the choices of electrodes are significant. Usually, lanthanides are well known to yield easily alloys or compounds with Cd, Bi and Al. In the case of Al, only cathodes for the electrochemical studies of Sm [4] and Dy [5] elements have already been reported. Moreover, the intermetallic compounds (Al3Sm, Al2Sm and Al3Dy) have been obtained by reaction of the reducing ion with Al substrate. However, the formation potential of Al-Sm and Al-Dy intermetallic compounds is close by the reports, which is disadvantageous to separation. We have studied the separation of SmCl3-DyCl3 system by potentiostatic electrolysis in KCl-LiCl-MgCl2 molten salts. However, galvanostatic electrolysis is often used in industry. Therefore, we study the separation of SmCl3 and DyCl3 by galvanostatic electrolysis in LiCl-KCl melts at magnesium electrodes. Fig. 1a shows the cyclic voltammogram at 1 mV s-1 in LiCl-KCl-SmCl3 solution on a Mg electrode at 773 K. It is found that the cyclic voltammogram in LiCl-KCl-SmCl3 melts is same to that in black LiCl-KCl melts at Mg electrodes. The cathodic signals Ac observed at approximately -2.28 V can be ascribed to the formation of Mg-Li alloys [6]. In the reverse scan, the anodic peaks Aa are detected, corresponding to the dissolution of Li from a layer of Mg-Li alloys. The sharp increase of cathodic current signal Bc and the corresponding anodic current peak Ba are associated with the reduction of Li(I) ions and dissolution of Li metal, respectively. It shows the formation potential of Mg-Sm intermetallic compound is more negative than that of Li metal. Fig. 1b shows the cyclic voltammogram at 1 mV s-1 in LiCl-KCl-DyCl3 solution on a Mg electrode at 773 K. Except for the redox signals Ac/Aa ascribed to the formation and dissolution of Mg-Li alloys, a part of new redox peaks Cc/Ca are detected and corresponding to the formation and dissolution of Mg-Dy alloys. Based on the result of cyclic voltammograms in Fig. 1, the current density (-0.02, -0.03 and -0.04 A cm-2) is used to study the separation of SmCl3 and DyCl3 by galvanostatic electrolysis on Mg electrodes. Table 1 shows the results of ICP-AES analysis of deposits after galvanostatic electrolysis experiments for 1 h. We think the separation of SmCl3 and DyCl3 can be achieved under -0.03 A cm-2 by galvanostatic electrolysis. 106
  • Table 1 The ICP analysis of deposits at Mg electrodes (S = 1.76 cm2) from LiCl-KCl-DyCl3-SmCl3 melts with different current density for 1 h at 773 K. Current density / A cm-2 -0.02 -0.03 -0.04 Content in deposits / wt. % Dy Sm Li Mg 4.2 10.9 12.6 0 0 2.7 1.9 1.9 2.1 Bal. Bal. Bal. Figure 1 Cyclic voltammograms: (a) in LiCl-KCl-SmCl3 (0.062 mol L-1) melts; (b) in LiCl-KCl-DyCl3 (0.062 mol L-1) melts. Working electrodes: Mg (S = 1.76 cm2); sweep rate: 1 mV s-1; temperature: 773 K. [1] L.J. Ott, R.N. Morris, J. Nucl. Mater., 371 (2007) 314. [2] H.S. Kim, C.Y. Joung, B.H. Lee, J.Y. Oh, Y.H. Koo, P. Heimgartner, J. Nucl. Mater., 378 (2008) 98-104. [3] H.P. Nawada, K. Fukuda, J. Phys. Chem. Solids, 66 (2005) 647-651. [4] Y. Castrillejo, P. Fernándeza, J. Medina, P. Hernández, E. Barrado, Electrochim. Acta, 56 (2011) 8638-8644. [5] Y. Castrillejo, M.R. Bermejo, A.I. Barrado, R. Pardo, E. Barrado, A.M. Martínez, Electrochimi. Acta, 50 (2005) 2047-2057. [6] A.M. Martínez, B. Børesen, G.M. Haarberg, Y. Castrillejo, R. Tunold, J. Appl. Electrochem., 34 (2004) 1271-1278. 107
  • Electrochemical extraction of holmium in LiCl-KCl-HoCl3 melts on a nickel electrode Tingting Sun, Wei Han*, Mei Li, Bin Liu, Pengyuan Sun Key Laboratory of Superlight Materials and Surface Technology, Ministry of Education, College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China * Corresponding author: weihan@hrbeu.edu.cn. The reprocessing of nuclear fuels today aims primarily to recycle lanthanides and actinides. Main objectives for future reactor systems are an effective fuel utilization and waste minimization through recycling of all lanthanides and actinides [1]. At present, a new field, the use of molten salts media for pyrochemical extraction as a promising option in the nuclear fuel, has been developed [2]. Since the deposition potentials of lanthanides and actinides are close to each other, the solid active electrodes or liquid cathodes are introduced to extract lanthanides and actinides from spent fuel [3]. In order to extract holmium from the bulk of fission products, the electrochemical behavior of holmium ions in LiCl-KCl melts was studied. Fig. 1(a) shows the voltammograms of LiCl-KCl eutectic melts before and after the addition of HoCl3 on a W electrode at 1023 K. The dotted curve represents the voltammogram before the addition of HoCl3. The signals A/A′ are corresponding to the reduction of Li (I) ions and dissolution of Li metal, respectively. In the solid curve in Fig. 1(a), except the peaks A/A′, the peaks B/B′ are associated with the deposition and subsequent reoxidation of holmium in LiCl-KCl eutectic melts. Fig. 1(b) shows the voltammograms of LiCl-KCl-HoCl3 melts on a Ni electrode at 1023 K. Besides the redox peaks A/A′ and B/B′, the observed peaks C/C′, D/D′ and E/E′ are attributed to the formation of Ni-Ho intermetallic compounds. Fig. 2 shows open-circuit polarization plateaus on Ni working electrode obtained after the polarization for 30 s at -2.3 V (vs. Ag/AgCl). Diffusion of holmium deposited on a Ni substrate results in the successive formation of Ni-Ho compounds. According to the Ho-Ni phase diagram, holmium can forms eight Ni-Ho intermetallic compounds with nickle. The first potential plateau B at about -2.06 V (vs. Ag/AgCl) is attributed to the dissolution of Ho(0). The plateaus C, D and E at about -1.87 V, -1.72 V and -1.01 V (vs. Ag/AgCl), respectively, indicate the formation of three kinds of Ni-Ho intermetallic compounds. Based on the results of cyclic voltammogram and open-circuit potentiogram, the Ni-Ho alloy compounds were prepared by potentiostatic electrolysis on a Ni electrode in the LiCl-KCl-HoCl3 melts at 1023 K. The Ni-Ho alloy compounds, Ni5Ho and Ni2Ho were characterized by XRD in Fig 3. The experimental results indicate that Ho can be extracted from the molten salts by Ni active electrode. Fig. 1 (a) Typical CVs before (dotted line) and after (solid line) the addition of 2.50×10-4 mol cm-3 HoCl3 in the LiCl-KCl melts on a W electrode (S=0.322 cm2) at 1023 K. Scan rate: 0.1 V s-1. (b) Typical CVs before (dotted line) and after (solid line) the addition of 2.50×10-4 mol cm-3 HoCl3 in the LiCl-KCl melts on a Ni electrode (S=0.322 cm2) at 1023 K. Scan rate: 0.1 V s-1. 108
  • Fig. 2 Open-circuit potential transient curve in LiCl-KCl-HoCl3 2.50×10-4 mol cm-3) melts on a Ni electrode (S=0.322 cm2) at -2.3 V for 30 s at 1023 K. Fig. 3 XRD patterns for the deposits obtained by potentiostatic electrolysis in LiCl-KCl-HoCl3 (2.50×10-4 mol cm-3) melt on Ni electrodes at 1023 K: (a) at -1.8 V for 10 h; (b) at -2.2 V for 10 h. [1] L. Cassayre, C. Caravaca, R. Jardinc, R. Malmbeck, P. Masset, E. Mendes, J. Serp, P. Soucek, J.-P. Glatz, J Nucl. Mater., 378 (2008) 79-85. [2] Y. Castrillejo, M.R. Bermejo, E. Barrado, Electrochim. Acta, 51 (10) (2006) 1941. [3] M.R. Bermejo, E. Barrado, A.M. Martínez, J Electroanal. Chem., 617 (2008) 85. 109
  • Electrochemical behaviors of REs in FLINAK eutectics Long Dewu*, Huang Wei, Jiang Feng, Tian Lifang, She Changfeng, Zhen Haiyang, Li Qingnuan Shanghai Institute of Applied Physics (SINAP), Chinese Academy of Sciences (CAS), Shanghai 201800 China * Corresponding author: longdewu@sinap.ac.cn. Tel: 86-21-3919 4677 Chinese Academy of Sciences had launched the “Thorium (Th)-based Molten Salt Reactor Nuclear Energy System (TMSR)” program in the beginning of 2011. Nowadays, a research group at SINAP is working on the R&D of pyroprocessing flowsheet for the treatment of the spent fuel from thorium-based molten reactor. Electrochemical separation method, one of the promising techniques, is being development. The electrochemical separation is based on the differences of redox potentials in the various species, which is strongly influenced by the physic-chemical properties of molten salt (MS) media. As the period of developing of Electrometallurgical Treatment (EMT) in America and Dimitrovgrad Dry Process (DDP) in Russia, a lot of endeavors had been paid on the study of electrochemical properties of actinides (Ans) and lanthanides (Lns) in chloride-based eutectic salts. However, the electrochemical data of such elements in fluoride-based eutectic salts are scarce. In order to acquire the fundamental data of REs in fluoride-based eutectic salts, the initial work of pyroprocessing research group at SINAP is focused on the study of the electrochemical behaviors of REs in LiF-NaF-KF (46.5-11.5-42 mol%, FLINAK) molten salts. Meanwhile, the preliminary experiments on the electrochemical deposition of REs from the FLINAK salts have been carried out. The results indicate that it is possible to extract Gd, Y and Nd from FLINAK eutectic salts because their electrode reactions are the reduction of trivalent cation to metal. The collective products at the cathode had been characterized by ICP-AES and SEM/EDS. The results show that these deposited products are REs-riched compounds, implying that direct electrochemical extraction of Gd, Y, and Nd is achievable. However, directly electro-deposition of Sm and Eu seems to be impossible since the reduction steps from cation to metal are out-range of the electro-stability of salt media. Therefore, alternative method has to be developed to extract Sm and Eu, such as using a reactive cathode is helpful. The formation of alloys with Sm and Eu at the cathode is significantly positive-shifted their reduction potential and therefore the extraction process would be achieved. 110
  • Electrochemical Behavior of Cerium and Electrodeposition of Al–Li–Ce Alloys from Molten Chlorides Zhang Menga, Han Weib*, Zhang Milinb, Li Meib, Xue Yuna, Yan Yongdeb, Zhang Zhijiana Fundamental Science on Nuclear Safety and Simulation Technology Laboratory; bKey Laboratory of Superlight Materials and Surface Technology, Ministry of Education, Harbin Engineering University, Harbin 150001, China * Corresponding author: weih@hrbeu.edu.cn a A study of the electrochemical behavior of CeCl 3 in LiCl–KCl molten salt at a Mo electrode or an Al electrode was investigated to ascertain if CeCl 3 is a suitable surrogate for PuCl3 in the development of nuclear fuel cycle technologies. The reaction mechanism and transport parameters of electroactive species are determined by electrochemical techniques. The results show that electrochemical reduction of Ce(III) at a Mo electrode in LiCl–KCl melts occurred in a reaction step with an exchange of three electrons. A voltammogram with a different scan rate in LiCl–KCl containing 0.125 M CeCl 3 shows that the deposition/dissolution reaction of Ce(III) ions is not completely reversible. The diffusion coefficient (D) of Ce(III) ions is estimated in the temperature range 833~923K by three methods. Fig. 1 compares the D values obtained in the present work with those obtained at the narrower testing temperatures of Iizuka[1], Castrillejo[2], Lantelme[3] and Pesic[4]. From an equation the calculated activation energy for diffusion of Ce(III) is found to be 32.6 kJ·mol-1. The expression for the Gibbs free energies of formation is equal to -1027.9+0.178T (K) kJ·mol-1. Furthermore, the standard rate constant is estimated to be (4.9~9.7) × 10-3 cm·s-1 by Nicholson in the scan rates range 10 ~ 80 mV·s–1 on the basis of cyclic voltammetry. The electroreduction of Ce(III) ions at an Al electrode in LiCl–KCl melts is also studied by cyclic voltammetry and open circuit chronopotentiometry in the temperature range of 668~742 K. The redox potential of Ce(III)/Ce at an Al electrode is observed at the more positive potentials values than those at an inert electrode. This potential shifts due to the formation of intermetallic compound with Al electrode. Al-Ce alloys are prepared in LiCl–KCl–CeCl 3 melts at 742K by potentiostatic electrolysis at an Al electrode. X-ray diffraction suggests that AlCe and AlCe 3 were formed in Al–Li–Ce alloys. A layer about 28µm was evenly coated the Al electrode in Fig.2. [1] M. Iizuka, J. Electrochem. Soc., 84 (1998). [2] Y. Castrillejo, M. Bermejo, R. Pardo and A. Martinez, J. Electroanal. Chem., 124 (2002). [3] F. Lantelme, T. Cartailler, Y. Berghoute and M. Hamdani., J. Electrochem. Soc., C604, (2001). [4] K. C. Marsden and B. Pesic, J. Electrochem. Soc., F111 (2011). Fig. 1. Variation of diffusion coefficient of CeCl3 with temperature in LiCl-KCl eutectic Fig. 2. Micrograph of the cross section of an Al electrode after electrolysis. 111
  • Electrochemical extraction of thulium in LiCl–KCl melt containing TmCl3 at liquid Zn electrodes Xing Lia, YongDe Yana,b, *, Yun Xuea,b, HaoTanga, DeBin Jia, Wei Hana, MiLin Zhanga,*, ZhiJian Zhangb a Key Laboratory of Superlight Materials and Surface Technology, Ministry of Education, College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China b Key Discipline Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin 150001, China * Corresponding author: y5d2006@hrbeu.edu.cn,zhangmilin@hrbeu.edu.cn. With the aggravating of energy shortage and environmental pollution, the exploitation and utilization of nuclear energy for electricity production are more and more emphasized by many nations. However, the widespread application of nuclear energy has produced a large quantity nuclear waste, which pose potential hazard to offspring. Therefore, the disposal of nuclear waste has attracted much attention. In the recent decades, the developed strategies for future nuclear reactor systems is to develop a closed fuel cycle including an effective fuel utilization and waste minimization through reprocessing of all actinides, especially the long-lived radionuclides [1]. For this reason, partitioning and transmutation (P&T) of long-lived fission products and minor actinides represents nowadays a promising method to dispose nuclear waste, which can recycle or destruct of the hazardous radionuclides into less hazardous or shorter lived elements [2]. However, before this transmutation, it is required to separate the minor actinides (MAs) from other fission products (FPs), especially the rare earth fission products (REE). Molten salt, in particular molten chlorides and fluorides, has been widely applied for the treatment and recovery of REE owing to its high thermal and radiation stability. Cyclic voltammograms for the redox reaction of thulium are shown in Fig.1. Curve 1 shows cyclic voltammograms obtained in LiCl–KCl–TmCl 3 melt on a W electrode at 723 K at 0.1 V/s. As can be seen, the reduction of Tm(III) ions into Tm metal takes place in two consecutive steps. The reduction of Tm(II) to Tm occurs at about –2.15 V. Curves 2 and 3 elucidate cyclic voltammograms obtained in LiCl–KCl melt before and after addition of TmCl 3 (2 wt.%) at liquid Zn electrode. Before the addition of TmCl 3 , at the beginning of scan, a huge anodic current corresponding to the anodic dissolution of liquid Zn electrode material is observed. The cathodic current at about –2.0 V is related to the underpotential deposition of Li(I) on a liquid Zn electrode forming a Zn–Li alloy. At more positive potential, the corresponding anodic current is related to the dissolution of the Zn–Li alloy. Here, the large anodic current at about –0.8 V and the huge cathodic current at about –2.0 V limit the potential window when liquid Zn is used as working electrode. The similar curves were obtained on liquid Cd and Bi electrodes in LiCl–KCl melt. After the addition of TmCl 3 , one couple of signals B/B′ at between Zn and Li reduction/oxidation was present, corresponding to the deposition/dissolution of Zn–Tm intermetallic compound. Considering the large surface area of liquid Zn electrode, the formed intermetallic compound should be Zn-rich Zn–Tm alloy. Fig. 2 shows the XRD pattern of the deposit. Zn and Tm2 Zn 17 phases were identified. However, Zn phase is predominant in the deposit which originated from the Zn electrode. 112
  • 3 0.2 B' 2 1 j/Acm-2 0.0 B -0.2 1 2 3 -0.4 -2.4 -2.0 W electrode liquid Zn electrode liquid Zn electrode -1.6 + -1.2 -0.8 E / V vs Ag /Ag Fig. 1. Curve1: Cyclic voltammograms obtained in LiCl–KCl–TmCl 3 (2 wt.%) melt on a W electrode (S=0.32 cm-2) at 723 K at 0.1 V/s; Curve 2, 3: Cyclic voltammograms obtained in LiCl–KCl melt before (curve 2) and after (curve 3) addition of TmCl 3 (2 wt.%) at liquid Zn electrode (S=1.38 cm-2) at 723 K at 0.1 V/s. 12000 Intensity/a.u. ♣ Zn ∆ Tm2Zn17 ♣ 10000 8000 ♣ 6000 4000 2000 ∆ 0 20 30 ∆ ♣ ∆ ♣ ∆ ∆ 40 ∆ 50 2θ/deg. 60 ♣ ♣∆ 70 ♣ ♣ 80 Fig. 2. The XRD pattern of the deposit obtained by galvanostatic electrolysis at –200 mA cm-2 for 6 h in LiCl–KCl–TmCl 3 (2 wt.%) melt at liquid Zn electrode at 753 K . [1] P. Souč ek, R. Malmbeck, E. Mendes, C. Nourry, D. Sedmidubský, J.-P. Glatz, J. Nucl. Mater. 394(2009)26. [2] M.R. Bermejo, J. Gómez, J. Medina, A.M. Martínez, Y. Castrillejo, J. Electroanal. Chem. 588 (2006)253 113
  • Electrochemical behavior of erbium and aluminum in the LiCl-KCl eutectic Kui Liu a, b, Ya-Lan Liua,d,Li-Yong Yuana, Xiu-Liang Zhaob, Hui Hed, Guo-An Yed, Yu-Liang Zhaoa, Zhi-Fang Chaia, c, Wei-Qun Shia* a Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049, China b School of Nuclear Science&Technology, University of South China, HengYang 421000 China c School of Radiological&Interdisciplinary Sciences, Soochow University, Suzhou 215123, China d Division of Radiochemistry, China Institute of Atomic Energy, Beijing 102413, China *Corresponding author: shiwq@ihep.ac.cn Rare earth metals and their alloys have been widely applied as advanced functional materials in numerous industrial fields due to their specific functional properties such as excellent magnetic[1] and optical [2] properties. Using molten salts as a reaction media, the electrowinning of high purity rare earth metals and synthesis of rare earth alloys can be well fulfilled[3, 4]. Another important field involving lanthanides (Ln) and molten salts is the so-called pyrochemical reprocessing of spent nuclear fuels, aiming at the partitioning and transmutation (P&T) strategy, from which the inventory of highly radioactive long half-lived minor actinides (MA) and fission products (FP) can be significantly incinerated to reduce their long term risk towards environment. This work presents an electrochemical study of Er3+ and Al3+ in the LiCl-KCl-AlCl3-Er2O3 melts at 773K. Gibbs energy calculation shows that AlCl3 can favorably chloridize Er2O3 and release Er3+ ions under this condition. Cyclic voltammetry (Fig.1), square wave voltammetry and chronopotentiometry were applied using a molybdenum electrode to investigate the co-reduction behavior of Er3+ and Al3+. The mechanisms and the Gibbs free energies of forming the biphasic coexisting Er-Al alloys were also investigated by open circuit chronopotentiometry. A series of redox signals corresponding to different kinds of Er-Al alloys were revealed. Potentiostatic and gavanostastic electrolysis (Fig.2) were both conducted at different on an aluminium electrode to prepare the Er-Al alloys. The obtained deposits were characterized by SEM-EDS and XRD. A layer of ErAl3 was obtained by potentiostatic electrolysis, whereas two kinds of alloys of ErAl3 and ErAl2 were formed by galvanostatic electrolysis. [1] S.A. Thomas, G.M. Tsoi, L.E. Wenger, Y.K. Vohra, S.T. Weir, Journal of Applied Physics, 111 (2012). [2] R.G. Elliman, A.R. Wilkinson, T.H. Kim, P.K. Sekhar, S. Bhansali, Journal of Applied Physics, 103 (2008). [3] M.R. Bermejo, F. de la Rosa, E. Barrado, Y. Castrillejo, Journal of Electroanalytical Chemistry, 603 (2007) 81-95. [4] H. Konishi, T. Nohira, Y. Ito, Electrochimica Acta, 47 (2002) 3533-3539. 114
  • Fig.1. CVs of AlCl3(2wt% )–Er2O3(1.5wt% ) at various switching potentials on the molybdenum electrode in the LiCl–KCl eutectic; Scan rates:100mV/s. Fig.2. SEM and XRD (inset) analyses of the cathodic deposits obtained in LiCl–KCl– AlCl3(2.5 wt.%)–Er2O3(1.5 wt.%) melts on an aluminium electrode at 773K. This work was supported by the Major Research Plan “Breeding and Transmutation of Nuclear Fuel in Advanced Nuclear Fission Energy System” of the Natural Science Foundation of China (Grants 91226201) and the "Strategic Priority Research program" of the Chinese Academy of Sciences (Grants.XDA03010401, XDA03010403 and XDA03010404). Tel/Fax: 86-10-88233968 http://shiweiqun.weebly.com 115
  • Electrochemical extraction of samarium from LiCl-KCl melt by forming Sm-Zn alloys Yan-Lan Liua,b, Li-Yong Yuana, Kui-Liua, Guo-An Ye*b, Mi-Lin Zhangc, Hui Heb, Hong-bin Tangb, Ru-Shan Linb, Zhi-Fang Chaia ,Wei-Qun Shi*a a Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049, China b Department of Radiochemistry, China Institute of Atomic Energy, Beijing, 102413, China c Key Laboratory of Superlight Materials and Surface Technology, Ministry of Education, College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China * Corresponding author: yeguoan@ciae.ac.cn; shiwq@ihep.ac.cn.   Zinc, with the low melting and boiling point, has several advantages when served as the molten cathode[1-3]. The present work is devoted to samarium extraction by forming Sm-Zn alloys in LiCl-KCl melt. The co-reduction of Sm(III) and Zn(II) ions on the Mo electrode and the under potential deposition of Sm(III) ions on the liquid Zn electrode were investigated in LiCl-KCl molten salt at 873 K. Cyclic voltammetry, square wave voltammetry and open-circuit chronopotentiometry techniques were employed to study the electrochemical behaviors of Sm(III), Zn(II) and Sm-Zn alloy formation. The results showed that on the inert Mo electrode the electro-reduction of Sm(III) took place in only one soluble-soluble electrochemical step Sm(III)/Sm(II), and the electrochemical step Sm(II)/Sm(0) was not observed within the electrochemical window. In contrast, the deposition potential of Sm(II)/Sm(0) shifted to the positive direction by forming SmxZny intermetallic compounds in both the co-reduction and under potential deposition process. In the co-reduction process, the signals of forming six kinds of SmxZny intermetallic compounds were observed in the high ZnCl2 containing melt, and the formation of SmxZny at about -1.56 V is dominant. In the under potential deposition process, the dominant deposition process is also at about -1.6 V. The formation of Sm-Zn alloys were carried out by potentiostatic electrolysis on both the Mo electrode and liquid Zn electrode. All the Sm-Zn alloys were characterized by scanning electron microscopy (SEM) with energy dispersive spectrometry (EDS), X-ray diffraction (XRD) and inductively coupled plasma atomic emission spectrometer (ICP-AES). The XRD results showed that the SmxZny compound prepared on the Mo electrode in the high ZnCl2 containing melt was SmZn12, while the SmxZny compound prepared on the liquid Zn electrode was predominantly Sm2Zn17. The SEM image of the products obtained by potentiostatic electrolysis on Mo electrode were shown in Fig. 1. [1] E.W. Murbach, W.N. Hansen, Ind Eng Chem 1959,51, 177-178. [2] T. Toda, T. Maruyama, K. Moritani, H. Moriyama, H. Hayashi, J. Nucl. Sci. Technol., 2009,46(1), 18–25. [3] A.V. Kovalevskii, V.A. Lebedev, I.F. Nichkov, S.P. Raspopin, Zhurnal Fizicheskoi Khimii 1974, 48(10), 2509-2512. SmZn12 Zn Fig.1.The SEM image of the Sm-Zn alloys obtained by potentiostatic electrolysis on the Mo electrode in LiCl-KCl-SmCl3 (2.0 wt.%)-ZnCl2 (5.0 wt.%) melt deposition at -1.6 V for 3h. Grants:XDA03010401,XDA03010403,XDA0301040 4(Chinese Academy of Sciences), Grants:91226201(Natural Science Foundation of China) http://shiweiqun.weebly.com 116
  • Molecular Dynamics Simulation of Molten LiF-ThF4 Salt Systems Jian-Biao Liu, Xin Chen, Chao-Fei Xu, W. H. Eugen Schwarz and Jun Li* Theoretical and Computational Chemistry Laboratory, Department of Chemistry, Tsinghua University, Beijing 100084, China * Corresponding author: junli@mail.tsinghua.edu.cn Molten fluorides are of interest because of their efficient use in different industrial processes, in particular in metallurgy and in nuclear energy applications such as the pyrochemical treatment of nuclear waste or the molten salt reactor (MSR). The local structure and transport properties of LiFThF 4 mixtures at different concentrations and temperatures were studied by using molecular dynamics simulations that include Born repulsion, Coulomb interaction, dispersion and dipole polarization effects. We have calculated the density, the self-diffusion coefficient, the electrical conductivity and the viscosity for this system. The radial distribution function and the variation of the coordination number of Th4+, of the number of bridging F atoms and of the lifetime of the first solvation shell are also discussed. [1] Salanne, M.; Rotenberg, B.; Jahn, S.; Vuilleumier, R.; Simon, C.; Madden, P. A. Theor. Chem. Acc. 2012, 131, 1143. [2] Dewan, L. C.; Simon, C.; Madden, P. A.; Hobbs, L. W.; Salanne, M. J. Nucl. Mater. 2013, 434, 322-327. [3] Pauvert, O.; Salanne, M.; Zanghi, D.; Simon, C.; Reguer, S.; Thiaudiere, D.; Okamoto, Y.; Matsuura, H.; Bessada, C. J. Phys. Chem. B 2011, 115, 9160-9167. Fig. 1. Snapshot of an instantaneous configuration from the simulation of LiF-ThF4 mixture. 117
  • Session 7: Innovative Materials and Separation 118
  • Study on proton beam irradiation of ionic liquid Yinyong Ao1, Jing Peng1, Shuojue Wang1, Jianyong Wang2, Ziqiang Zhao2, Maolin Zhai1* 1 Beijing National Laboratory for Molecular Sciences, Radiochemistry and Radiation Chemistry Key Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, China 2 State Key Laboratory of Nuclear Physics and Technology, School of Physics, Peking University, Beijing 100871, PR Chin * Corresponding author: mlzhai@pku.edu.cn. Due to their attractive physical and chemical properties, the room-temperature ionic liquids (RTILs) are considered as promising solvents for liquid-liquid extraction of radioactive isotopes in the nuclear fuel recycle.[1,2] In the practical application of extraction solvents, they must be resistant to very strong irradiation, in particular, γ-irradiation and α-irradiation. Therefore, studies on the radiation effect of RTILs and their extraction ability of metal ions are very important for the assessing application feasibility in the reprocessing of spent nuclear fuel. Recently, we reported that the radiolytic products of [C 4 mim][NTf 2 ] under γ-irradiation were HF, H 2 SO 3 , CF 3 SO 2 NH 2 , CF 3 SOOH and CF 3 SO 2 OH, and these radiolytic products play important roles during the extraction of Sr2+ from water to irradiated [C 4 mim][NTf 2 ].[3,4] In order to study the Fig 1. The equipment for the proton beam irradiation. radiation effect of RTILs under α-irradiation, we designed special irradiation equipment (Fig 1) for simulating the α-irradiation with proton beam. The ionic liquid samples were irradiated at room temperature with 2 MeV proton beam and the stopping and range of proton beam in ionic liquids was calculated using SRIM 2012 Monte Carlo code. The separation of water-soluble radiolytic products from organic phase was conducted by contacting 0.5 mL of irradiated sample with 0.5 mL of deuterium oxide (D 2 O) for about 10 min in a vibrating mixer, followed by centrifuging to ensure that the phases were fully contacted and then separated. The aqueous phase from the wash of irradiated [C 4 mim][NTf 2 ] was analyzed by Micro-FTIR, 19F NMR and high-resolution ESI-MS. As indicated in Fig 2, the 19F NMR of unirradiated sample shows a simplex peak at -78.78 ppm, which is assigned to NTf 2 - based on previous literature[5], but that of irradiated sample shows many signals of radiolytic products after proton irradiation. The main radiolytic products were identified and the exact chemical structure was confirmed by using high resolution ESI-MS, Micro-FTIR and 19F NMR spectrum of standard compound. Based on our previous work[4], the peak at -129.59 and -164.02 ppm can be assigned to the signal of SiF 6 2- and HF, respectively. The signal at around -87.7 ppm is assigned to CF 3 SOOH based on the Micro-FTIR analysis (OH band at 3300 - 3500 cm-1 and 1600 - 1700 cm-1) and 19F NMR of CF 3 SOOH standard compound. In addition, CF 3 SO 2 NH 2 was identified by high-solution ESI-MS analysis (m/z = 147.9079, the exact m/z of CF 3 SO 2 NH- is 147.9080), Micro-FTIR (NH 2 band at 954 cm-1) and 19F NMR (-79.33 ppm); CF 3 SO 2 OH was confirmed by high-solution ESI-MS analysis (m/z = 148.9517, the exact m/z of CF 3 SO 2 O- is 148.9520) and 19F NMR (-75.6 ppm). The above results indicates that the radiolytic products of [C 4 mim][NTf 2 ] under proton beam irradiation are similar to those under 119
  • γ-irradiation, and the radiation effect of [C 4 mim][NTf 2 ] under He+ beam is underway. Fig 2 The 19F NMR spectra of [C 4 mim][NTf 2 ] (unirradiated sample (a) and irradiated sample (b)). References [1] Binnemans, K. Lanthanides and actinides in ionic liquids. Chem. Rev. 2007, 107, 2592-2614. [2] Xu, C.; Yuan, L. Y.; Shen, X. H.; Zhai, M. L. Efficient removal of caesium ions from aqueous solution using a calix crown ether in ionic liquids: mechanism and radiation effect. Dalton Trans. 2010, 39, 3897-3902. [3] Yuan, L. Y; Xu, C.; Peng, J.; Xu, L.; Zhai, M. L; Li, J. Q; Wei, G. S; Shen, X. Identification of the radiolytic product of hydrophobic ionic liquid [C 4 mim][NTf 2 ] during removal of Sr2+ from aqueous solution. Dalton Trans. 2009, 7873-7875. [4] Ao, Y. Y; Peng, J.; Yuan, L. Y; Cui, Z. P; Li, C.; Li, J. Q; Zhai, M. L. Identification of radiolytic products of [C 4 mim][NTf 2 ] and their effects on the Sr2+ extraction.Dalton Transactions 2013, 42, 4299-4305. [5] Berthon, L.; Nikitenko, S. I.; Bisel, I.; Berthon, C.; Faucon, M.; Saucerotte, B.; Zorz, N.; Moisy, P. Influence of gamma irradiation on hydrophobic room-temperature ionic liquids [BuMeIm]PF 6 and [BuMeIm](CF 3 SO 2 ) 2 N. Dalton Trans. 2006, 2526-2534. 120
  • Surface modification of carbon nanomaterials and their application in radionuclide pollution cleanup Xiangke Wang Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031, China. Email: xkwang@ipp.ac.cn Keywords: plasma technique, carbon nanomaterials, water pollution cleanup Abstract Heavy metal ions and radionuclides in the natural environmental have attracted world attention because of their hazardous to the environment and human health. Thereby, the elimination of the pollutants is crucial and different kinds of natural materials such as oxides and minerals and manmade nanomaterials such as carbon nanotubes have been investigated extensively to evaluation their potential application in the removal of pollutants. Although nanomaterials have high sorption capacity in the adsorption of pollutants from aqueous solution, the non-selectivity of the nanomaterials restricted the application of nanomaterials in real applications. Herein, we reported a novel method, plasma technique, to modify functional groups on the surface of nanomaterials to make them highly selectively removal the pollutants from large volumes of aqueous solutions. The plasma technique is an environmental friendly method to graft functional groups on the surface of materials without destroy the structure of the materials. 1. Zhao GX et al.,Advanced Materials 2011, 23, 3959-3963. 2. Yang SB et al., Environmental Science & Technology. 2011, 45(8), 3621- 3627. 3. Zhao GX et al., Environmental Science & Technology. 2011, 45, 10454- 10462. 4. Sun YB et al., Environmental Science & Technology,2012, 46,6020-6027. 5. Wang Q et al., Chemistry-An Asian Journal. 2012, 8, 225-231 121
  • Extraction uranium from aqueous solution with malonamide into ionic liquid Liu, Ziyi; Li,Wenkui; Wu; wangsuo; Shen, Yinglin* a Radiochemistry Laboratory, Lanzhou University, Lanzhou 730000, P. R. China Corresponding author: shenyl@lzu.edu.cn The removal of uranium from various matrices, in addition to its continuing relevance to nuclear materials processing, remains a separation of importance, not only in industrial applications but also in energy and environmental problems [1]. Up to now, many separation and preconcentration techniques for metal ions have been developed, such as the liquid–liquid solvent extraction [2] and the solid phase extraction (SPE) [3]. Room-temperature ionic liquids (ILs) have recently attracted much attention as environmentally favorable solvents because of their practically negligible vapor pressure[4]. Malonamide is the promising extractants for trivalent actinide ions. Advantages of malonamide ligands include ease of synthesis, milder stripping requirements, and chemical compositions consisting of fully incinerable elements (i.e., no P or S present) that adhere to the CHON principle[5]. Its solvent extraction properties for the actinides, lanthanides and fission products have been reported in detail using n-dodecane as the diluent. The major one is associated with a large volume of n-dodecane used as the process diluent, which tends to form a third phase with nitric acid and generates a large amount of the secondary waste. In addition, potential health hazard and flammability risk are also associated with it. It is, therefore, imperative to look for an alternative technique where the secondary waste volume can be reduced. In this work, Extraction of the uranyl, UO 2 2+, from nitric acid solution was carried out using N,N,N′,N′-tetrabutylmalonamide as a extractant into an ionic liquids (ILs) of 1-alkyl-3-methylimidazolium bis(trifluoromethane)sulfonamide ([C n mim][NTf2 ], n=4,6,8), which was shown as Figure 1. The distribution coefficient for UO 2 2+ ion decreases with increasing nitric acid concentration in the aqueous phase due to the protonation of malonamide molecule. At low HNO 3 concentration, ILs containing butyl groups functionalized with imidazolium cation show higher values of D compared to octyl groups functionalized with imidazolium cation. The extraction mechanism was deduced by the slope analysis and extraction tests. The UO 2 2+ 122
  • extraction most likely occurs by a mechanism involving a cation-exchange. In addition UO 2 2+ ions form 1:2 complexes with malonamide in ILs at lower acidity, and 1:1 complexes in ILs at higher acidity. 25 20 D 15 [C4mim]NTf2 [C6mim]NTf2 [C8mim]NTf2 10 5 0 0 1 2 3 4 CHNO3 M 5 6 Fig. 1 Dependency of UO 2 2+ distribution ratio D on HNO 3 concentration at constant(20mM) malonamide concentration in ILs. 2+ Fig. 2 Plots logD vs logC, [UO 2 ] = 0.4 mM. References [1] G.J. Lumetta, K.L. Nash, S.B. Clark, J.I. Friese (Eds.), Separations for the Nuclear Fuel Cycle in the 21st Century, American Chemical Society, Washington, DC,2006, pp. 21–40. [2] S.K. Sahu, V. Chakravortty, M.L.P. Reddy, T.R. Ramamohan, The synergistic extraction of thorium(IV) and uranium(VI) with mixtures of 3-phenyl-4- benzoyl-5isoxazolone and crown ethers, Talanta 51 (2000) 523–530. [3] C. Cuillerdier, C. Musikas, L. Nigond, Diamides as actinide extractants for various waste treatments, Sep. Sci. Technol. 28 (1993) 155–175. 123
  • [4] (1) a. Han, X. X.; Armstrong, D. W. Acc. Chem. Res. 2007, 40, 1079–1086. b. Rogers, R. D.; Voth, G. A. Acc. Chem. Res. 2007, 40,1077-1078. c. Sun, P.; Armstrong, D. W. Anal. Chim. Acta 2010, 661, 1–16. [5] C. Cuillerdier, C. Musikas, L. Nigond, Diamides as actinide extractants for various waste treatments, Sep. Sci. Technol. 28 (1993) 155–175. 124
  • Extraction of uranium(VI) and thorium(IV) ions from the aqueous phase into an ionic liquid by 4-oxaheptanediamides a Peng Rena, Ze-Yi Yana,* Radiochemistry Laboratory, School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000, PR China * Corresponding author: yanzeyi@lzu.edu.cn The exploration of various actinide separation routes has been driven by the need for technology to successfully process the large quantity of high-level nuclear wastes. Although a variety of separation technologies are available, the efficacy of liquid/liquid separations is virtually unparalleled. Due to some advantages, such as the ability to subtly or dramatically change system conditions, associated rapid kinetics, and the large selection of organic solvents available, liquid/liquidseparation are regarded as a mainstay to target the separation of actinides and/ortheir fission products from aqueous solutions in the practical REDOX and the PUREX processes. However, in traditional liquid/liquid system mentioned above, the need to use volatile or semi-volatile organic compounds or the subsequent generation of organic compounds, introduces risk associated with the solvent’s toxic and flammable nature. In addition to the obvious health problems associated with organic diluents, strict environmental policies ensure a high cost of spent solvent disposal. The development of novel solvents for use in separations provides an opportunity to extend the concepts and practices of ‘green’ technology in the nuclear fuel cycle. Ionic liquids are promising solvent alternatives in liquid/liquid extraction processes. In the past decade, extraction and separation of actinides and lanthanides from nuclearwaste using ionic liquids has already been widely reportedas they show high stabilityunder α- and γ-irradiation and enhance safety towards criticality. O O N O O N O N O TOHA O N TEHA O O O O N N N O N THHA TBHA Figure 1. Extractants used in this study. In this study, we envisioned to N,N,N',N'-tetraoctyl-4-oxaheptanediamide (TOHA) 125 synthesize tridentate ligand and analogous diamides
  • (N,N,N',N'-tetraethyl-4-oxaheptanediamide,TEHA; N,N,N',N'-tetrabutyl-4-oxaheptanediamide, TBHA; N,N,N',N'-tetrahexyl-4-oxaheptanediamide, THHA), and investigated their extraction behavior to actinides Figure 1). These 4-oxaheptanediamide (OHDA) ligands were assumed to ( be associated with the metal ion to form stable two six-memberedrings, and drastically increase the affinity with metal ions. In the present study, we described the synthesis and characterization of different types of 4-oxaheptanediamide derivatives, and their extraction properties for uranium(VI) and thorium(IV) ions from nitric acid solution in various ionic liquids(Figure 2). 4.0 TEHA TBHA 3.0 THHA log D TOHA 2.0 1.0 0.0 -4.7 -4.6 -4.5 -4.4 -4.3 log c -4.2 -4.1 -4 Figure 2. Distribution ratios of U(VI) as a function of concentrations of ligands at 25±1 °C, [U(VI)] = 0.01mol/L,[HNO 3 ] = 0.01mol/L, [C 4 mim+]PF- as diluent. [1] E.P. Horwitz, D.G. Kalina, H. Diamond, G.F. Vandegrift,W.W. Schulz, Solv. Extr. Ion Exch. 3, 75 (1985). [2] A. M. Rozen, Z. I. Nikolotova, N. A. Kartasheva,Radiokhimiya, 28,407 (1986). [3] Z. Kolarik, U. Müllich, F. Gassner, Solvent Extr. Ion Exch.,17, 23 (1999). [4]A.E. Visser, M.P. Jensen, I. Laszak, K.L. Nash, G.R. Choppin, R.D. Rogers, Inorg. Chem. 42, 2197 (2003). [5] K. Nakashima, F.Kubota, T. Maruyama, M. Goto, Ind.Eng. Chem. Res. 44, 4368 (2005). [6] A.E. Visser, R.P. Swatloski, W.M. Reichert, S.T. Griffin, R.D. Rogers, Ind. Eng. Chem. Res.39, 3596 (2000). [7] M.L. Dietz, J.A. Dzielawa, Chem.Commun. 2124 (2001). [8] A.E. Visser, R.D. Rogers, J. Solid State Chem.171, 109 (2003). [9] A.V. Mudring, S. Tang, Eur.J. Inorg. Chem. 2569 (2010). 126
  • Radiation Effect on EuIII Extraction Ability of BTPhen/ILs System Hanyang Zhoua, Yinyong Aoa, Jie Yuana, Jing Penga,*, Jiuqiang Lia and Maolin Zhaia,* a Beijing National Laboratory for Molecular Sciences (BNLMS), Radiochemistry and Radiation Chemistry Key Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, China * Corresponding authors: Tel/Fax: +86-10-62757193, E-mail: jpeng@pku.edu.cn; mlzhai@pku.edu.cn Minor actinide (MA) transmutation by fast neutron, leading to residues with less radiotoxic and shorter half-life nuclides, is very efficient and widely anticipated in advanced nuclear fuel cycle, but it is required to separate them from lanthanides(Ln) before transmutation, due to the existence of Ln in the spent nuclear fuel and much higher neutron capture section of Ln[1,2]. 2,9- bis (5,5,8,8- tetramethyl -5,6,7,8- benzo [1,2,4] triazin-3-yl)- 1,10 –phenanthroline (BTPhen) is an extractant that has been reported to be efficient in traditional volatile diluents (such as 1-octanol) for Am/Eu separation, which is typical MA/Ln separation[3]. In our investigation, the extraction ability and radiation stability of EuIII by BTPhen employing ionic liquids (ILs) as diluents was evaluated for the first time. Fig. 1 demonstrates that ILs with a shorter alkyl side chain in the imidazolium cations is favorable for the extraction of EuIII at neutral, due to its lower viscosity and higher cation exchange activity. However, in the presence of nitric acid, the distribution ratio of Eu (DEu) decreases dramatically, and the length of alkyl chain of cations in ILs has slight effect on the extraction, since the neutral species extraction became the dominant mechanism instead of cation exchange. Nevertheless, compared with BTPhen/1-octanol system, BTPhen/[C2mim][NTf2] system still has higher extraction property. It was found that DEu for BTPhen/1-octanol system decreased 3 orders at dose of 200 kGy, compared with unirradiated BTPhen/1-octanol indicating the serious radiolysis of BTPhen induced by the radiolysis of 1-octanol. However, ILs could protect the BTPhen and the extraction property of BTPhen/[C2mim][NTf2] system had no obvious change at same dose. Therefore, BTPhen/[C2mim][NTf2] extraction system has better extraction ability and higher radiation stability than BTPhen/1-octanol system, which is expected to be a promising Ln-MA separation agent in the advanced nuclear fuel cycle. [1] J. Magill, V. Berthou, D. Haas, J. Galy, R. Schenkel, H. W. Wiese, G. Heusener, J. Tommasi and G. Youinou, Nucl. Energ.-J. Br. Nucl. 42, 263 (2003). [2] J. N. Mathur, M. S. Murali and K. L. Nash, Solvent Extr. Ion. Exc. 19, 357 (2001). [3] F. W. Lewis, L. M. Harwood, M. J. Hudson, M. G. B. Drew, J. F. Desreux, G. Vidick, N. Bouslimani, G. Modolo, A. Wilden, M. Sypula, T.-H. Vu and J.-P. Simonin, J. Am. Chem. Soc. 133, 13093 (2011). 127
  • Fig.1. DEu of BTPhen in different [Cnmim][NTf2] ionic liquids (■: [C2mim][NTf2]; ●: [C4mim][NTf2]; ▲: [C8mim][NTf2]. Equilibration time: 1 h). 128
  • --Template-Separation of Uranyl Species Using Task-specific Ionic Liquid, [Hbet][Tf2N] Takahiro MORIa, Kotoe SASAKIb, Tomoya SUZUKIa, Tsuyoshi ARAIb, Koichiro TAKAOa, Yasuhisa IKEDAa a Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-34, Ookayama, Meguro-ku, Tokyo, 152-8550 Japan b Department of Materials Science, Shibaura Institute of Technology 3-7-5 Toyosu, Koto-ku, Tokyo 135-8548, Japan * Corresponding author: yikeda@nr.titech.ac.jp Task-specific ionic liquids (TSILs) with functional groups designed to give particular properties or reactivity are paid attention as attractive alternatives to the conventional organic solvents. As one of TSILs, protonated betaine bis(trifluoromethylsulfonyl)imide ([Hbet][Tf2N]) has interesting properties, that is, this IL can dissolve metal oxides such as Ln2O3 (Ln = lanthanoid), UO3, PbO, CuO, etc., and become miscible with water with an increase in temperature[1,2]. Hence, [Hbet][Tf2N] is expected to be used as media for treating radioactive wastes contaminated by oxides of radioactive nuclides. In the present study, in order to investigate feasibility of the above proposal, we have examined the extraction behavior of uranyl species and some other metal species from aqueous phase to [Hbet][Tf2N] one and compared with that in the [TMPA][Tf2N] (TMPA = trimethylpropylammonium) syetem(see Fig. 1). The [Hbet][Tf2N] was synthesized according to the literature method and [TMPA][Tf2N] was purchased from Kanto Chemical Co., Inc. The extraction experiments were performed by mixing IL solution (2 ml) with aqueous HNO3 solution (2 ml, [U(VI)] = 20 mM) dissolved Hbet UO2(NO3)2·6H2O for 1 h at 25 ºC, followed by centrifugation. The concentrations of [U(VI)] in aqueous phase were measured ICP-AES. Figure 2 shows plots of extractability (E, %) of U(VI) against [HNO3]. TMPA Tf2N The U(VI) species can be extracted to [Hbet][Tf2N] phase in the low Cation Anion acid region, while in the [TMPA][Tf2N] sysyem U(VI) species are not extracted to IL phase. This suggests that the –COOH group of Hbet Fig. 1. Structures of ILs. contributes to the extraction of uranyl species. In the low acidic region, the proton of –COOH group of Hbet is dissociated and hence the uranyl complexes coordinated with COO- should be formed. The resulting uranyl species should be extracted to the [Hbet][Tf2N] phase, because the uranyl species in [Hbet][Tf2N] are known to be exist as [UO2(bet)3]-. Furthermore, we confirmed that Na(I), Al(III), Ni(II), and Co(II) are not extracted from aqueous phase to [Hbet][Tf2N] one. From these result, it is expected that the uranyl species can be extracted and separated selectively and that [Hbet][Tf2N] solution should be used as the decontamination media. [1] P. Nockemann, B. Thijs, S. Pittois, J. Thoen, C. Glorieux, K.V. Hecke, L.V. Meervelt, B. Kirchner, K. Binnemans, J. Phys. Chem. 110, 20978 (2006). [2]P. Nockemann, R.V. Deun, B. Thijs, D. Huys, E. Vanecht, K.V. Hecks, L.V. Meervlt, K. Binnemans, Inorg. Chem. 49, 3351 (2010). 129 ― ― Fig. 2. Plots of extractability of U(VI) vs. [HNO3] in the extraction systems used [Hbet][Tf2N] and [TMPA][Tf2N] as the extraction media.
  • Dissolution of UO2 in the system of [Imim][FeCl4]-DMSO Aining YAOa, Taiwei Chua* a Beijing National Laboratory for Molecular Sciences, Radiochemistry and Radiation Chemistry Key Laboratory of Fundamental Science, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, China * Corresponding author: twchu@pku.edu.cn Ionic liquids (ILs) can offer low volatility, wide liquid range, thermal stability, and enhanced criticality safety, and these properties can make the IL-based nuclear fuel cycle reprocessing safer than that based on volatile organic solvents. Imidazolium-based ILs show relatively high radiation resistance attributed to the presence of aromatic ring which can absorb radiation energy and can relax nondissociatively1. They are more thermally stable than the alkyl ammonium ionic liquids because they are resistant to ring fission during thermal rearrangements2.Fe-containing ionic liquids have been reported by substituting FeCl4- for Cl-, and the incorporation of FeCl4- ion may change the physicochemical properties of the original IL, e.g., hydrophilic 1-butyl-3-methylimidazolium chloride [Bmim]Cl turning into hydrophobic [Bmim]FeCl43. Moreover, FeCl4- brings high oxidation ability, and nano-sized conducting polymers can be prepared without additional oxidants4. A recent report has shown that Imidazolium-based Fe-containing ionic liquids (ILs) can directly dissolve UO2 in the presence of their corresponding imidazolium chlorides without additional oxidants5. With the help of Cl-, UO2 can be successfully changed to [UO2Cl4]2-. Besides chloridion, DMSO can also be used as uranyl’s coordination agent. In this study, imidazolium-based Fe-containing ionic liquids [Imim]FeCl4 are used to dissolve UO2 in the presence of DMSO. With thoroughly stirring, the black UO2 powder gradually dissolves in the ILs, turning the ILs phase from cloudy to clear. The dissolution process was measured using ICP-AES to obtain the dissolved amount of uranium at different time as shown in Fig. 1, and % U dissolved means the percentage of dissolved uranium species compared with the initial addition of uranium. At the beginning, the dissolved uranium percentage increases exponentially, which is an experimental indication of first-order kinetics. A plot of ln[(C∞-C)/C∞] versus time is shown in Fig. 2, where C is the concentration of dissolved uranium at time t and C∞ is taken as the concentration at 900 min. There is a linear relationship between ln[(C∞-C)/C∞] and t, indicating that the initial dissolution appears to follow a pseudo first-order kinetics. The slope of the line is 0.023 min-1, which may be regarded as the rate constant of the initial pseudo first-order dissolution process of UO2 in 1ml [Bdmim]FeCl4 containing 0.75 ml DMSO at 120 oC. In the medium of imidazolium-based Fe-containing ILs containing their corresponding imidazolium chlorides, uranyl(VI) species has been reported to precipitate in the form of [imidazolium]2[UO2Cl4]5. Different from Cl- anion, the precipitant in the medium of DMSO has a different form. Both of elemental analysis and IR spectrum show that the precipitant does not contain imidazolium cations, and mass spectrum and Raman spectrum show that FeCl4- exists in the anion part. The exact structure of the precipitant is currently in progress. [1] K. Binnemans, Chem. Rev. 107, 2592 (2007). [2] W. H. Awad, J. W. Gilman, M. Nyden, R. H. Harris Jr, T. E. Sutto, J. Callahan, P. C. Trulove, H. C. DeLong and D. M. Fox, Thermochim. Acta 409, 3 (2004). 130
  • [3] K. Bica and P. Gaertner, Org. Lett. 8, 33 (2006). [4] J.-Y. Kim, J.-T. Kim, E.-A. Song, Y.-K. Min and H.-o. Hamaguchi, Macromolecules 41, 2886 (2008). [5] A. Yao and T. Chu, Dalton Trans. 42, 8413 (2013). Fig. 1. % U dissolved in 1ml [Imim]FeCl4 IL and 0.75 ml DMSO at 120oC as a function of time. (Imim = Emim+, Bmim+, Bdmim+) 131 Fig. 2. Plot of ln[(C∞-C)/C∞] versus time for the dissolution of UO2 in 1 ml [Bdmim]FeCl4 containing 0.75 ml DMSO at 120 oC.
  • Influence of solvent structural variations on the isoBu-BTP/[C n mim][NTf2 ] extracting system during Eu(III)/Dy(III) extraction Guolong Maa, Shenggu Maa, Weijin Yuana, Long Zhaoa*, YuezhouWeia a Nuclear Chemical Engineering Laboratory, School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, P. R. China * Corresponding author: Tel/Fax +86-21-34207654; E-mail: ryuuchou@sjtu.edu.cn. 1-alkyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide (abbreviated hereafter as [C n mim][NTf 2 ]), imidazolium-type ionic liquids (ILs), used as green solvents now have become a research hot spot in the field of separation of meal ions by solvent extraction [1]. In current work, the liquid-liquid extraction of lanthanides from nitric acid solution using [C n mim][NTf 2 ] based extracting system was investigated. The novel extracting system is made up of 2,6-bis(5,6-dibutyl-1,2,4-triazin-3-yl) pyridine (isoBu-BTP), which acted as extractant, can separate lanthanides and minor actinides effectively in nitric acid solution, and a series of 1-alkyl-3-methylimidazolium-based ILs which acted as extraction medium, show a great solubility against isoBu-BTP[2]. The current work was focused on the extraction and separation of two typical Ln(III): Eu(III) and Dy(III), as the [C n mim][NTf 2 ] varied in alkyl chain length. Scheme 1 The Structure of [C n mim][NTf 2 ] and isoBu-BTP It was found that the alkyl chain length plays a remarkable role in determining the dominant mode of extraction via Ln(III) transfer in the isoBu-BTP/[C n mim][NTf 2 ](n=2, 4, 6, 8) extracting system. Fig.1 demonstates that ILs with a shorter alkyl chain in the imidazolium cations extracted the Ln(III) extremely quickly. The extraction of Dy(III) in isoBu-BTP /[C 2 mim][NTf 2 ] system is about 1000 times faster than that occurred in isoBu-BTP/ [C 6 mim][NTf 2 ] system. It was also found that the length of alkyl chain can affect the extraction efficiency and extraction selectivity in current work. With increasing the alkyl chain length, the hydrophobicity becomes increasingly significant, and the extraction mechanism of lanthanide metal from aqueous solutions into ionic liquids in the presence of isoBu-BTP is consistent with a shift from cation exchange to extraction of the neutral complex. Current work can gain some new insight into isoBu-BTP/[C n mim][NTf 2 ] extracting system. In addition to that obvious fundamental significance, the results obtained in current work may also have practical applicaition in the design of improved methods for Ln(III)/An(III) separation in advanced nuclear fuel cycle. Therefore further research on this topic is needed to be done. 132
  • 0 500 Time (min) 1000 1500 2000 2500 3000 80 80 60 [C2mim]+ [C mim]+ 40 4 [C6mim]+ [C mim]+ 20 8 60 40 E (%) 100 E (%) 100 20 0 0 0 5 10 15 Time (min) 20 25 30 Fig. 1. Dy (Ⅲ) extraction kinetics by isoBu-BTP/ [C n mim][NTf 2 ] (n=2,4,6,8). Aqueous phase: [Dy3+]=2mM, HNO 3 concentration=0.01M. Extracting phase:[Iso-Bu-BTP]=22mM using [C n mim][NTf 2 ]. (O:V=1:1, T=25℃) [1] Sun X., Luo H., Dai S., Chemcial Reviews. 112: 2100-2128(2012). [2] Panak, P.J. and A. Geist, Chemical reviews. 113(2): 1199-1236(2013). 133
  • Extraction of several rare-earth metal ions using isoBu-BTP/[C2mim][NTf2] system Shenggu Ma, Guolong Ma, Weijin Yuan, Long Zhao*, YuezhouWei Nuclear Chemical Engineering Laboratory, School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, P. R. China * Corresponding author: Tel/Fax +86-21-34207654; E-mail: ryuuchou@sjtu.edu.cn. Hydrophobic ionic liquids are promising extraction media in solvent extraction separation of metal ions[1]. In this work, the extraction behavior of several rare-earth metals (Y, Nd, Eu, Dy) from nitric acid solution was studied using an novel extracting system based on hydrophobic ionic liquid. The novel extracting system is made up of 2,6-bis(5,6-dibutyl-1,2,4-triazin-3-yl) pyridine (isoBu-BTP), which acted as extractant, can separate lanthanides and minor actinides effectively in nitric acid solution , and 1-ethyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([C2 mim][NTf2]),which acted as extraction medium, shows a great solubility against isoBu-BTP [2]. The influence of the concentration of nitric acid, the contact time, the concentration of extractant and the stripping condition was investigated in detail. H F CH2CH3 N N H3C N N F F N C C F N N N H N H F O O F S S N O O Scheme 1 The structure of isoBu-BTP (left) and [C2mim][NTf2] (right). Fig.1 demonstrates that the isoBu-BTP/[C2 mim][NTf2] extracting system is favorable to extract rare-earth metals at low acidity condition (<0.1M), and back-extract at high acidity condition (>3M). In addition, Dy was found to be selectively extracted in current extracting system. The extraction mechanism for the extraction of Dy3+ was discussed. A lg-lg plot of the DDy values versus isoBu-BTP concentration gave a straight line with a slope of 2.4 in current research (see Fig.2). It is suggested that most Dy from aqueous phase was extracted by BTP, which can formed a 3:1 molar complex with Dy(III), while a small quantity of Dy was extracted by ionic liquid itself. The cation exchange and coextraction, which induced form ionic liquid, functioned to extraction and resulted into such stoichiometry mentioned above. Lastly, the comparison between isoBu-BTP/[C2 mim][NTf2] system and isoBu-BTP/SiO2-P system was also performed, even they have different mechanism, but the study was focused on the different separation effect between these two novel separation systems. It was found that the ionic liquid system is favorable to lower acidity conditions (<0.1M) while SiO2-P system is favorable to higher acidity conditions (>1M) (see Fig.3).This opposite dependence of acidity may help us to design some new separation process for hydrometallurgy industry or advanced nuclear fuel cycle application. 134
  • 0.8 100 Y Nd Eu Dy 80 0.6 0.4 lg[D] E(%) 60 40 20 0.2 0.0 -0.2 0 -0.4 0.0 0.5 1.0 1.5 2.0 2.5 3.0 [HNO3] /M 1.3 1.4 1.5 1.6 lg[BTP]/mM Fig.2. Extraction of Dy3+by isoBu-BTP in [C2 mim][NTf2], with distribution ratios as a function of isoBu-BTP concentration. Fig.1. Effect of concentration of HNO3 on extraction efficiency of Y3+, Nd3+, Eu3+, 100 80 80 60 60 40 40 20 20 0.0 0.5 1.0 1.5 2.0 [HNO3 ]/ M 2.5 3.0 Y Eu Nd Dy Nd Dy Y Eu extraction efficiency (%) adsorption efficiency (%) 100 0 1.7 0 Fig.3. The adsorption efficiency of isoBu-BTP/SiO2-P system (red) and the extraction efficiency of isoBu-BTP/[C2 mim][NTf2] system (blue). [1] Dai, S., Y. Ju, and C. J. Chem. Soc., Dalton Trans. (8): 1201-1202(1999). [2] Panak, P.J. and A. Geist, Chemical reviews. 113(2): 1199-1236(2013). 135
  • Electrodeposition of Rh(III) and Pd(II) from 1-Ethyl-3-Methylimidazolium Trifluoroacetate Shuai Gua, Xinpeng Wanga, Yuezhou Weia,* (School of Nuclear Science and Engineering, Shanghai Jiao Tong University 800 Dongchuan Road, Shanghai 200240, China Telephone: 86-21-34205684 ) * Corresponding author: yzwei@sjtu.edu.cn. The fission products of Uranium(U) and Plutonium(Pu) covered from Zinc to Lutecium(Lu) in periodic table of the elements, some of them are rare earth metals, such as Rhodium(Rh) and Palladium(Pd). This paper studied the electrodeposition of Rh and Pd from 1-Ethyl-3-Methylimidazolium Trifluoroacetate. The Standard electrode potential of Rh(III), and Pd(II) differ by 0.2V approximately, also the ionic liquid performs better than aqueous in electrodeposition, which may make the electrodeposition of Rh and Pd respectively becomes possible. Materials used in this work were analytical pure. The experiments were carried out at 80℃ in a conventional three-electrode system. The working electrode was a glassy carbon(GC) electrode (Φ = 2 mm) with an apparent exposed area of 3.14 mm2. Its potential was monitored against a KCl-saturated Ag/AgCl reference electrode (0.197 V vs. SHE). A 10×10×0.3 mm platinum plate electrode was used as the counter electrode. All electrode potentials reported in this work are referred to the potential of the KCl-saturated Ag/AgCl reference electrode. Prior to each experiment, surface pretreatment of the working and counter electrode were performed by hand polishing the electrode surface with 1.0, 0.3, 0.05 micro Gamma-Alpha alumina powder in a sequence to a mirror finish and then immersed in 3M HCl, acetone, methanol and distilled water in ultrasonic cleaner for 20 minutes, respectively, to remove any surface impurity.The electrolyte was consisted of analytical grade of 5 mM Pd(NO3)2 or/and 5mM RhCl3. Fig. 1 shows the Cyclic voltammetry(CV) of GC electrode in 1-Ethyl-3-Methylimidazolium Trifluoroacetate solution containing 5mM Pd(NO3)2 at the scan rate of 10mV/s with the potential scan starting from 0.70V to 0.30V and then reverse to 1.20V. As we can see from Fig. 1 that during the cathodic scan there was only one peak (a) at 0.52V, also at the reverse scan there was one peak (b) at 0.89V which was the stripping peak of Palladium. A couple of sensitive and reversible redox peaks were obtained from Fig. 1. While Fig. 2 shows the Cyclic voltammetry(CV) of GC electrode in 1-Ethyl-3-Methylimidazolium Trifluoroacetate solution containing 5mM RhCl3 at the scan rate of 10mV/s with the potential scan starting from 0.40V to -0.20V and then reverse to 0.50V. We can also observe two peaks, peak (A) at -0.03V was the cathodic peak of Rh and peak (B) at 0.020V was the anodic peak. Fig. 3 shows the Cyclic voltammetry(CV) of GC electrode in 1-Ethyl-3-Methylimidazolium 136
  • Trifluoroacetate solution containing 5mM Pd(NO3)2 and 5mM RhCl3 at the scan rate of 10mV/s with the potential scan starting from 0.40V to -0.20V and then reverse to 0.50V. Fig. 1 Cyclic Voltammetry of Pd(II) in ionic liquid solution Fig. 2 Cyclic Voltammetry of Rh(III) in ionic liquid solution We can see four peaks in Fig. 4, peak (C) was the cathodic peak of Pd(II) at 0.30V, which is a little lower than the solution without Rh(III); peak (D) was the cathodic peak of Rh(III) at -0.040V; peak (E) and (F) were the anodic peak of Rh and Pd. As a result of the difference between peak (C) and (D) are 0.34V, which is big enough for the electrodeposition of Pd(II) and Rh(III) respectively. Fig. 3 Cyclic Voltammetry of Pd(II) and Rh(III) in ionic liquid solution 137
  • Adsorption of Thorium on Magnetic Multi-walled Carbon Nanotube 1 Peng Liu1, Jing Wang1, Wei Qi1, Wangsuo Wu1* Radiochemistry Laboratory, Lanzhou University,Lanzhou,Gansu, China, 730000 * Corresponding author: wuws@lzu.edu.cn With the world energy crisis deepening, nuclear power has been developed vigorously, thorium as a kind of potential nuclear fuel has been used widely. But the radioactive thorium maybe come into the biosphere and enriched in organism through the water cycle, so the research on purification of thorium pollutants in water environment has important significance. Carbon nanotubes have excellent physiochemical property and versatile applications. However, CNTs suffer from separation inconvenience, in order to further improve their properties and functions, iron oxide nanoparticles were aggregately coated on the surface of CNTs. The influences of pH, contact time, solid-liquid ratio, temperature and C60(C(COOH)2)n on Th(IV) adsorption onto the magnetic multi-walled carbon nanotubes (MMWCNTs) were studied by using batch technique. The dynamic process showed that the adsorption of Th(IV) onto MMWCNTs could be in equilibrium for about 40h and matched the pseudo-second-order kinetics model well. The adsorption process was independent on the ionic strength. The adsorption of Th(IV) on MMWCNTs was significantly dependent on pH values, the adsorption rate increased markedly at pH 3.0–5.0 and then maintained the highest level with the increasing of pH values. The effects and behavior of Th(IV) adsorption onto MMWCNTs were studied when the C60(C(COOH)2)n was added. C60(C(COOH)2)n could affected the adsorption of Th(IV) onto MMWCNTs obviously, at low pH C60(C(COOH)2)n enhanced the adsorption, but C60(C(COOH)2)n blocked the adsorption at higher pH. The Langmuir, Freundlich and Dubini-Radushkevich models were used to describe the adsorption characteristics of adsorbent in water and wastewater treatment, respectively. It can be concluded from the constants that the Langmuir and D-R models fit the experimental data better than by the Freundlich model. The adsorption isotherms indicated that the adsorption was endothermic,and the adsorption of Th(IV) on MMWCNTs occurs mainly by surface complexation. And the adsorption was irreversible. 0.0014 0.0012 298K 323K 100 80 Sorption(%) Cs(mol/g) 0.0010 0.0008 0.0006 0.0004 60 40 20 None C60(COOH)n C60(COOH)n=75mg/L C60(COOH)n=125mg/L C60(COOH)n=250mg/L 0.0002 0 0.0000 0.00000 0.00001 0.00002 0.00003 0.00004 0.00005 0.00006 0.00007 0 2 4 6 8 10 Ce(mol/L) pH Fig.1 Effect of different temperature of Th(IV) adsorption onto MMWCNTs Fig.2 Effect of C60(C(COOH)2)2 on Th(IV) adsorption onto MMWCNTs 138
  • A catechol-like phenolic ligand-functionalized hydrothermal carbon: one-pot synthesis, characterization and sorption behavior towards uranium Bo Li, Yin Tian , Xiaodan Yang, Juan Li, Chiyao Bai, Xiaoyu Yang, Shuang Zhang, Lijian Ma, Shoujian Li* College of Chemistry, Sichuan University, Key Laboratory of Radiation Physics & Technology (Sichuan University), Ministry of Education, Chengdu 610064, P. R. China. * Corresponding author: sjli000616@scu.edu.cn (S. Li). Uranium is one of the most important source materials for nuclear energy industry. Thus, separation and recovery of uranium from various uranium-containing aqueous systems are of great scientific and practical significance. In the present work, we proposed a new approach for preparing an efficient uranium-selective solid phase extractant (HTC-btg) and by choosing bayberry tannin and glyoxal via a simple, economic, and green one-pot hydrothermal synthesis. Same way was employed for preparation of another HTC-based material (hereafter named HTC-bt) using only bayberry tannin as starting material for comparison with HTC-btg. The results of characterization and analysis shown that after addition of glyoxal into only bayberry tannin-based hydrothermal reaction system, the as-synthesized HTC-btg displayed larger specific surface area and more than doubled surface phenolic hydroxyl groups. The sorption behavior of the sorbents towards uranium under various conditions was investigated in detail. The U(VI) sorption capacity reached up to 307.3 mg g−1 under the current experimental conditions. The selective sorption in a specially designed multi-ion solution containing 12 co-existing cations shown that the amount of uranium sorbed accounts for about 53% of the total sorption amount at pH 4.5. The designed scheme for the preparation of HTC-btg by hydrothermal synthesis process in this work is illustrated below. Scheme 1. Preparation of HTC-btg by hydrothermal synthesis process. The optimum conditions for preparation of HTC-btg were decided by orthogonal tests. BET results indicated that the specific surface area of HTC-btg increased to 207.04 m2 g−1, which is six times over that of HTC-bt. The pore volume also increased to 0.388 L g−1, which is eight times more than that of HTC-bt. Boehm titration [1] shown that the phenolic group content of HTC-btg increased significantly to 6.67 mmol g−1, more than double that of HTC-bt, and the carboxyl group contents in both THC-bt and HTC-btg remained almost the same(Table. 1), which indicated that bayberry tannin could formed polymeric 3D networks by glyoxal in the hydrothermal process. The result is consistent with data BET results. According to the results of competitive adsorption test (Fig.1), the U(VI) sorption amount of HTC-btg reached 0.42 mmol g−1 much higher than that of HTC-bt (0.28 mmol g−1) with the total sorption amount increased from 0.69 mmol g−1 for HTC-bt to 0.80 mmol g−1 for HTC-btg. Moreover, the sorption amounts of HTC-btg for main group metals (Ni2+, Zn2+, Co2+, Sr2+ etc.) is significantly lower than that of HTC-bt. The combination of these two factors makes the amount of uranium sorbed 139
  • accounts for about 53% of the total sorption amount significantly higher than that of HTC-bt (about 40%). The proposed strategy could be extended easily to obtain new HTC-based functional materials with as many of desired functional groups as is available, promising for wider range of applications, such as selective separation and recovery of valuable metals, wastewater treatment and catalytic reaction. Further study is being undertaken. [1] H.P Boehm, carbon. 40 145-149 (2002). Table 1. Content of carboxylic groups and phenol groups on the surface of HTC-bt and HTC-btg determined by Boehm titration. Sample Carboxyl group (mmol g−1) Phenolic group (mmol g−1) HTC-bt 3.79 ± 0.36 3.07 ± 0.66 HTC-btg 3.30 ± 0.88 6.67 ± 0.29 Fig. 1. Competitive sorption capacities of coexistent ions on HTC-bt and HTC-btg (c0 ≈ 0.5 mmol L−1, pH = 4.5, t = 120 min, V = 25 mL, T =298.15 K, and w = 10 mg). 140
  • A Simple Approach to Highly Efficient Uranium Selective Sorbent: Preparation and Performance of a Novel Amidoxime-functionalized Hydrothermal Carbon Xiaodan Yanga, Juan Lia, Jun Liub, Yin Tiana, Songbai Liua, Shoujian Lia,* Lijian Maa,* College of Chemistry, Sichuan University, Key Laboratory of Radiation Physics & Technology, Ministry of Education, Chengdu 610064, P. R. China. b Nuclear Power Institute of China, Yunke Pharmaceutical Corporation, Chengdu, 610041, P. R. China. * Corresponding author: sjli000616@scu.edu.cn (S. Li), ma.lj@163.com a Uranium is an important nuclear energy resource. Separation and recovery of uranium from various uranium-containing effluents are of great significance to sustainable development of nuclear energy, human health as well as environment protection [1]. This research is performed to explore an alternative approach for preparation of new hydrothermal carbon (HTC)-based materials with uranium-specific selective function. In the current study, two simple and common molecules, glyoxal and acrylonitrile, are chosen, for the first time, as starting materials to prepare a new amidoxime (AO)-functionalized HTC-based solid phase extractant (HTC-AO) via one-step hydrothermal process following a simple oximation procedure (as depicted in Scheme 1) [2,3]. Scheme 1. Schematic illustration of the preparation of HTC-gly, HTC-CN and HTC-AO. Results from characterizations using Boehm titrations, XPS and N 2 adsorption show that the as-synthesized product HTC-AO achieves the anticipated surface property and framework structure, i.e., low porosity frameworks (0.01 cm3 g-1) and intraparticle diffusion coefficient (k int = 0.042 mmol g−1 min−0.5), the overwhelming majority of amidoxime group (1.66 mmol g-1) and the minimal of undesired functional groups (typically carboxylic group: 0.07 mmol g-1; phenolic group: 0.38 mmol g-1; lactonic group: 0.01 mmol g-1) on the surface of HTC-AO, which could help to greatly improve uranium selectivity of the resulting material. The sorption behavior of U(VI) ions onto HTC-AO was investigated in detail using batch sorption experiments. Distinctively, a saturation U(VI) sorption capacity over that of all the uranium sorbents [2-6] reported previously is found to be 1021.6 mg g−1 at pH 4.5 in pure uranium solution. In addition, the selective sorption of uranium onto the HTC-AO was investigated in a multi-cation solution sample containing 12 competitive nuclide ions besides uranyl ion [7]. The results in Fig. 1 show that the sorption amount of HTC-AO towards U(VI) raise to 1.56 mmol g−1 (about 3 times more than that of HTC-gly and HTC-CN), with 2.44 mmol g−1 of total sorption capacity for all nuclide ions in the multi-cation solution. The uranium-selectivity (S U ) for HTC-AO is calculated up to 64.0%, about 2.5 times higher than that for HTC-gly and HTC-CN ( S U ≈ 27% for each). On the other hand, as shown in Fig. 1b for HTC-AO, the K d value reaches high up to nearly 5000 mL g−1 for U(VI) and fairly low (< 141
  • 300 mL g−1) for other coexistent ions. And with pH varying over the ranges 1.0 to 4.5 (Fig. 2), the selectivity towards uranium appears to be increasing at first and then decreasing slightly after reaching the peak value at approximately pH 2.5 (a so far unreported highest uranium selectivity of 81.6% with a sorption capacity of 268.9 mg g-1) in multi-ion solution (inset in Fig. 2). Fig. 1. (a) Competitive sorption capacities and the (b) K d of coexistent ions on HTC-gly, HTC-CN and HTC-AO (c 0 = 0.84 mmol L−1 for U(VI) and 1.0 mmol L−1 for other cations, pH = 4.5, t = 120 min, V = 25 mL, T = 298.15 K, and w = 10 mg). Fig. 2. Effect of pH on the sorption of U(VI) in multi-ion system (c 0 = 1.0 mmol L−1 for all cations, t = 120 min, V = 25 mL, T = 298.15 K, and w = 10 mg). The significant outcomes in this work may offer a practical strategy for solving the bottleneck problems of poorer selectivity and lower sorption capacity existing in common sorption materials, and demonstrate that the as-prepared HTC-AO possesses excellent ability for selective sorption of uranium especially in low pH solution and is considered to be a promising candidate for the separation of uranium from nuclear effluents as well as other related water sources. [1] G. J. Lumetta, K. L. Nash, S. B. Clark, Separations for the nuclear fuel cycle in the 21st century; American Chemical Society: Washington, DC, 2006. [2] S. Das, A. K. Pandey, A. A. Athawale, J. Phys. Chem. B., 113, 6328 (2009). [3] J. Górka, R. T. Mayes, L. Baggetto, J. Mater. Chem. A., 1, 3016 (2013). [4] P. J. Lebed, J. D. Savoie, J. Florek, Chem. Mater., 24, 4166 (2012). [5] M. J. Manos, M. G. Kanatzidis, J. Am. Chem. Soc., 134, 16441 (2012). [6] J. Qu, W. Li, C. Y. Cao, J. Mater. Chem., 22, 17222 (2012). [7] C. R. Preetha, J. M. Gladis, T. P. Rao, Environ. Sci. Technol., 40, 3070 (2006). 142
  • Amidoxime-Grafted Multiwalled Carbon Nanotubes by Plasma and its Application in the Removal of Uranium Yun Wanga,b, Jun Tanga* , Zexing Gua, Jiali Liao, Jijun Yang, Yuanyou Yang, Ning Liua a Key Laboratory of Radiation Physics and Technology (Sichuan University), Ministry of Education; Institute of Nuclear Science and Technology, Sichuan University, Chengdu 610064, China b Fundamental Science on Radioactive Geology and Exploration Technology Laboratory, East China Institute of Technology, NanChang, Jiangx , 330013,China *tangjun@scu.edu.cn Multiwalled carbon nanotubes (MWCNTs) have attracted multidisciplinary interest because of their special physicochemical properties. MWCNTs have been proven to possess excellent adsorption capacity to adsorb heavy metal ions from large volume of aqueous solutions. However, the property of less surface functional groups greatly restricts their application. Surface modification can significantly enhance the adsorption capacity after functional groups are introduced on adsorbent surfaces. Plasma induced grafting treatment are promising method to introduce functional groups on adsorbent surfaces without using large amount of chemicals and altering their bulk properties. Herein, a new amidoxime-functionalized (AO) multiwalled carbon nanotubes have been successfully prepared by using plasma techniques. The AO grafted MWCNTs (AO-g-MWCNTs) were characterized by Fourier transform infrared spectra (FT-IR), Raman spectra, Power X-ray diffraction (XRD), scanning electron microscopy (SEM), and Nitrogen-BET methods in detail. The application of AO-g-MWCNTs in the removal of uranium from aqueous solution was investigated through batch experiments, including the effect of pH, sorbent dosage, contact time, temperature, initial uranium concentration and ionic strength on uranium sorption. Sorption of U(VI) on the sorbent was pH-dependent. Sorption equilibrium was reached in 60 min. Distinctively, higher temperature was beneficial to the sorption of U(VI) in the range of 10-55℃, high ionic strength had almost no effect on the sorption, and the maximum U(VI) sorption capacity of 196mg g−1 was observed under the condition stested.The AO-g-MWCNTs have much higher sorption ability in the removal of uranium than raw MWCNTs and are a suitable material in the preconcentration and solidification of uranium from large volume of aqueous solutions. 143
  • Amino Functionalized MIL-101 Metal–Organic Frameworks (MOFs) for U(VI) Capture Zhi-QiangBaia,b, Cheng-Liang Xiaoa, Li-Yong Yuana,Zi-JieLia,LinWanga, Zhi-RongLiub, Yu-Liang Zhaoa, Zhi-Fang Chaia, Wei-Qun Shi*a a Key Laboratory of Nuclear Radiation and Nuclear Energy Technology and Key Laboratory for Biomedical Effects of Nanomaterials and Nanosafety, Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049, China; bSchool of Nuclear Engineering and Geophysics, East China Institute of Technology, Nanchang 330013, China. *Corresponding author: shiwq@ihep.ac.cn The development of nuclear power plant has a rapidly growing over the past decades and is predicted to be in a continuous increase in the future. However, due to thenuclear fuel cycle and mining operations in the nuclear power industry,more attentions must be paid to the human safety and environmental issues caused by the radionuclides [1,2].Metal-organic frameworks (MOFs) are a class of hybrid materials with increasing applications in many areas such as sorption[3], catalysis,gas separation and storage. In this work, the acidity-stable MIL-101(Cr) MOF was synthesizedand functionalized with amino groups by post-grafting. The synthesized MIL-101, MIL-101-NH 2 , andMIL-101-ED (ED: Ethanediamine), MIL-101-DETA (DETA: Diethylenetriamine) were characterized with XRD, FT-IR, N 2 sorption, SEM, and TGA. It is found that the amino groups were successfully grafted onto the MIL-101 substrate and the structures did not change. The sorption properties of these materials towardsU(VI) from acidic solution were investigated. The amino functionalized materials were shown to be highly efficient in capturing U(VI)compared to the original MIL-101. The sorption capacities of these materials for U(VI) wereMIL-101-DETA> MIL-101-ED > MIL-101-NH 2 > MIL-101 with ~90 mg g-1,~200 mg g-1and~300 mg g-1for MIL-101-NH 2 , MIL-101-ED, and MIL-101-DETA, respectively. This work provides a facile and efficient way to modify MIL-101 MOF and these materials can be appliedto capture uranium and other radionuclides from acidic media. [1] L. Y. Yuan, Y. L. Liu, W. Q. Shi, Y. L. Lv, J. H. Lan, Y. L. Zhao and Z. F. Chai, Dalton Trans., 40, 7446,(2011). [2] L. Y. Yuan, Y. L. Liu, W. Q. Shi, Z. J. Li, J. H. Lan, Y. X. Feng, Y. L. Zhao, Y. L. Yuan and Z. F. Chai, J. Mater. Chem., 22, 17019,(2012). [3] W Yang, Z-Q Bai, W-Q Shi, L-Y Yuan, T Tian, Z-F Chai, H Wang and Z-M Sun, Chem. Commun., 49, 10415, (2013). Fig.2The sorption isotherms of MIL-101 MOFs for U(VI) sorption. This work was supported by the Natural Science Foundation of China (Grants 91026007, 11205169, 11275219) and the "Strategic Priority Research program" of the Chinese Academy of Sciences (Grants. XDA030104). 144
  • A Novel Functionalized 2-D COF Materials: Synthesis and Application as Selective Solid-phase Extractant in Separation of Uranium Juan Li, Xiaodan Yang,Chiyao Bai, Shuang Zhang, Lijian Ma*, Shoujian Li* College of Chemistry, Sichuan University, Chengdu 610064, PR China * Corresponding author: sjli000616@163.com Uranium, as the most important nuclear energy resources, its separation, enrichment and recovery from nuclear industrial effluents, mine water, seawater and other liquid mediums have great significance in either effective utilisation of resources or environmental security[1]. Up to now, the separation and enrichment method of uranium mainly include liquid-liquid extraction, solid-phase extraction (SPE) and precipitation method etc.. In recent years, solid-phase extraction (SPE) has been given close attention for its more advantages including higher enrichment factors, low consumption of reagents (especially organic solvent), absence of emulsion formation, minimal secondary waste generation, flexibility and so on [2]. A solid phase extractant usually consist of two parts, the solid phase matrix and the functional component. Commonly used solid phase matrix mainly contain Al2O3[3], silicagel[4], diatomaceous earth[5], polymer resin[6], and carbonaceous materials[7]. Although these materials have their own characteristics, there are still inevitable shortcomings more or less, for practical use, such as nonresistant of acid and radiation, unsuitable for use in high temperature environment and less surface area. Therefore, much attention of researchers has been paid to explore novel solid-phase matrixes that can overcome all the aforementioned deficiencies. COFs (Covalent organic frameworks) were reported for the first time as an emerging class of crystalline porous polymers by O. M. Yaghi in 2005[8]. COFs can be categorized into either two- (2D) or three-dimensional (3D) ordered porous structure and be only composed of light elements (e.g. C、 H、O、B、Si) through strong covalent bonds (C-C, C-O, C-Si, B-O). Based on many outstanding advantages as light, porosity, high specific surface areas, high thermal stability, especially regular, orderly and size-adjustable pore structure, COFs played a huge role in the field of gas storage and separation, catalyst support, optical imaging, etc[9]. Further, these features also provide a possibility for COFs as a solid phase matrix material to apply in the field of the separation and enrichment of the metal ions. For the above-mentioned reasons, a new type of solid-phase extractant that consists of COF as a matrix was developed and applied to the separation and enrichment of uranium in simulated spent fuel solution. Further, the synthesized COF need to be modified through introduction of functional groups that can selectively coordinate UO22+ ion. For containing both coordinating atoms N and O, benzimidazole compounds are considered as good functional components with high selectivity and affinity for uranyl ion. In addition, the compounds can be obtained through a simple one-step reaction with readily available and cost-effective raw materials. In this work, we synthetised a new COF which is abundant in –COOHs by the way of reserving the corresponding functional groups in raw materials directly and appropriately controlling the reaction conditions (solvent, the ratio of raw materials, temperature, time, etc.). On the other hand, a novel benzimidazole compound, 2-(2,4-dihydroxyphenyl)-benzothiazole (HBI) was synthetised by a one-step reaction of o-phenylenediamine and 2,4-dihydroxy benzaldehyde. By the way, the HBI was demonstrated to indeed possess high selectivity and extraction rate for uranyl ion by liquid-liquid extraction experiments in the presence of multi-ions. Finally the thus-formed solid-phase extractant COF-HBI was obtained through esterification reaction of the –OH in HBI and the -COOH on surface of the COF (Scheme 1). The result of FTIR spectra and thermogravimetric analysis indicated that the COF-HBI possessed high thermal and radiation stability. The material was found to have fairly desirable selectivity and affinity for U(VI) ions through a series of static adsorption experiments in the 145
  • multi-ion and pure uranium solutions system (Fig.1.). Therefore, The COF-HBI is considered as a promising candidate in the application of the separation and recycling of uranium from uranium-bearing solution, such as nuclear fuel effluents, seawater, saline lake water etc. Meanwhile, it also offers more choices for the preparation of solid-phase extraction agent . Scheme 1 Schematic illustration of the preparation of COF-COOH and COF-HBI OH ClOC COCl N HO COCl 4mmol + NH2 COOH HN 50 mL toluene 15ml THF 45 ml EA 0 °C for 1h, RT for 24h COF reflux at 110 °C for 12h COF-COOH N H O N O OH COF COF-HBI NH2 3mmol Fig. 1. Competitive adsorption capacities of coexistent ions on COF-COOH and COF-HBI (C0≈0.5 mmol L−1, pH = 4.5, t = 120 min, V = 25 mL, T = 298.15 K, and w = 10 mg). [1] Y. Arai, C.C. Fuller, J. Colloid Interface Sci., 365, 268–274 (2012). [2] Z. Guo, Y. Li, S. Zhang, H. Niu, Z. Chen, J. Xu, J. Hazard. Mater.,192, 168–175 (2011). [3] J. Tashkhourian, L.M. Abdoluosofi, M. Pakniat, M. Montazerozohori, J. Hazard. Mater., 187, 75– 81 (2011). [4] S. Sadeghi, E. Sheikhzadeh, Microchim. Acta, 163, 313–320 (2008). [5] M. Sprynskyy, I. Kovalchuk, B. Buszewski, J. Hazard. Mater., 181, 700–707 (2010). [6] A. Rahmati, A. Ghaemi, M. Samadfam, Ann. Nucl. Energy, 39 ,42–48 (2012). [7] M. Ghaedi, S.Z. Amirabad, F. Marahel, S.N. Kokhdan, R. Sahraei, M. Nosrati and A. Daneshfar, Spectrochim. Acta, Part A, 83, 46–51 (2011) . [8] A. P. Cote, A. I. Benin, N. W. Ockwig, M. O'Keeffe, A. J. Matzger and O. M. Yaghi, Science,310, 1166–1170 (2005). [9]. Xiao Feng , Xuesong Ding and Donglin Jiang, Chem. Soc. Rev.,41, 6010-6022 (2012). 146
  • Comparation of Ce(IV) stripping rate from TBP and DBP Chen Zuoa, Taihong Yana, Weifang Zheng a,*, Siwei Shia a China Institute of Atomic Energy * Corresponding author: wfazh@ciae.ac.cn Batch experiments have been done to compare the Ce(IV) stripping rate from 5% tributyl phosphate and 5% dibutyl phthalate(a main radiolysis product of tributyl phosphate) in xylene diluent, and reductants of different extractability were tested in these experiments. The results show that, extractability of reductant have little effect on Ce(IV) stripping rate from 5%TBP- xylene, but reductant with higher distribution ratio significantly strip Ce(IV) faster from 5%DBP- xylene, as shown in Table 1. Due to the excellent performance of acetaldoxime in stripping Ce(IV) from DBP, another study was performed with this reductant in organic phase. The results show that, acetaldoxime can reduce Ce(IV) rapidly when the organic phase contain a certain mount of TBP, but can not reduce Ce(IV) effectively if there is only DBP, the cause of this phenomenon may be the poor extractability of acetaldoxime in DBP. According to the results, it can be assumed that, when TBP was used as extractant, TBP-Ce(IV) complex dissolved in aqueous phase can dissotiate into Ce4+, and then Ce4+ reacts with reductant, the main reaction area is aqueous phase, so the extractability of reductant doesn’t play an important role. But in the case of DBP which is a stronger complexant, DBP-Ce(IV) complex dissolved in aqueous phase might be far less than TBP-Ce(IV), and difficult to dissotiate into Ce4+ due to the possible chelated structure of DBP-Ce(IV) complex as in the case with DBP-Pu(IV) complex, the main reaction area is organic phase, so the extractability of reductant becomes very important in Ce(IV) stripping from DBP. Considering the similarities of Ce(IV) and Pu(IV), this might also explain why hydroxylamine strip Pu(IV) from DBP very slowly, and U(IV) which is extractable drastically improved the efficiency of Pu stripping[1,2]. Studies are currently under way to gain a better understanding of the real mechanism which is important in Pu(IV) reductant selecting. [1] Baron P, Boullis В, Germain M, et al. Proceedings of GLOBAL, (1993). [2] Kawaguchi Y, Morimoto K, Kitao T, et al. Nippon Genshiryoku Gakkai Wabun Rombunshi, 8(3), (2009). [3] Sun Xuemei[D]. Beijing: China Institute of Atomic Energy, (2012). [4] YE Guoan. Atomic Energy Science and Technology, 38(2),(2004). Table I. Time need for Ce(IV) stripping Distribution ratio of From 5%TBP reductant in 30%TBP[3,4] Hydroxylamine 1.16×10-4 ~10s Dimethylhydroxylamine 7.16×10-5 ~10s Acetaldoxime 0.03-0.1 (in 1B) ~10s Reductant From 5%DBP >30min ~2min ~20s Table II. Time need for Ce(IV) reduction by acetaldoxime in organic phase 5%DBP- xylene >3d 5%TBP- xylene ~1min 2.5%TBP-2.5%DBP- xylene ~2min 147
  • Impact of low molecular weight organic acids on uranium uptake and distribution in a variants of mustard (Brassica juncea var.tumida) Fangfang Qi 1, Liang Du 2, Xiaojie Feng 3, Zhongyong Cha 3, Bing Qin 3, Dingna Wang 1, Dong Zhang 2, Zhendong Fang 3, Yongdong Jin 1, Chuanqin Xia 1 1. College of Chemistry, Sichuan University, Chengdu 610064, China 2. Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900, China 3. Logistic Engineering University, Chongqing 401311, China Abstract: This study aimed at exploring the mechanism on natural low molecular weight organic acids (LMWOAs) effects on uranium (U) accumulation and distribution in plants. Citric acid, malic acid, or mixture of citric acid/malic acid/oxalic acid/ lactic acid were chosen as models, and Mustard (Brassica juncea var.tumida) growing in China was selected as test plant. Dynamics and kinetics of U desorption from U-contaminated soil by organic acids was tested, LMWOAs effects on U speciation in soils were studied by sequential extraction procedure. Then U accumulation kinetics in plants in four organic acids series was investigated. Finally, we fractioned plant tissues and assessed U distribution in roots, stems and leaves by usingmicroX-ray fluorescence (μ-XRF) mapping. These results indicated that organic acids could increase uranium leaching yields from 1% (control) to 46-57%, and change the fraction of different uranium speciation in soil, especially increase the carbonate and exchangeable fraction of uranium. U accumulation kinetics in plants showed that addition of citric acid or mixed acid into exposure solution could enhance root-to-shoot translocation, but inhibit U accumulation in roots. In addition, μ-XRF analysis observed that U was distributed in vascular tissues and cortex in plant roots and mainly in vascular tissues in stems and leaves in control, while mainly in vascular tissues in the whole plants in citric acid medium. These results suggested that U might be accumulated and transported through vascular tissue in plant with citric acid treatment. Thus, our observation would be helpful to properly develop efficient uranium phytoremediation. Keywords: Uranium;Low molecular weight organic acids; Accumulation; Distribution;μ-XRF;Phytoremediation B 22000 1.0 0.9 0.8 0.7 Rediues Organic Fe-Mn oxides Carbonate Exchangeable Percentage 0.6 0.5 0.4 0.3 0.2 0.1 0.0 control MA CA Mix Root uranium concentration (mg kg-1 d.w.) A 24000 C control CA MA Mix 20000 18000 16000 14000 12000 10000 8000 6000 4000 2000 0 1 2 3 5 7 9 14 Time (days) Fig.1 The impact of organic acids on U speciation in uranium-contaminated soils (Fig.1A),the kinetics of root uranium accumulation with organic acids addition(Fig.1B),error bars represent sample standard deviation (n=3),and a variants of mustard used in this study (Fig.1C). 148
  • Sorption of selenium(IV) on modified bentonites Hai Wanga,b, Tao Wu a*, Qing Zhenga, Jiang Chenc, Yao Lin Zhaob a Department of Chemistry, Huzhou Teachers College, Huzhou 313000, P. R. China b School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049, P. R. China c Huzhou Environmental Protection Monitoring center, Huzhou313000, P. R. China *corresponding author:twu@hutc.zj.cn Bentonite has been selected as the candidate of backfill material for nuclear waste repositories because of its large specific area, low hydraulic conductivity, low cost, high cation exchange capacity for most of radionuclides. However, it has a negligible sorption of anionic radionuclides like79Se(half-life 3.27×105 a) due to anion exclusion between negative charge surface of bentonite and anionic species. The objective was to modify the surface charge of bentonite and increase the sorption of anionic radionulides. Gao miaozi (GMZ) bentionite was modified to produce inorgano- and organo-bentonites for the removal of Se(IV) ions from aqueous solution. Inorgano-bentonites were modified with Fe, Al, Mg, La, FeOH and AlOH, whereas organo-bentoniteswere modified respectivelywithCetylpyridinium Chloride(CC-M), Cetylpyridinium ChlorideandSodium Stearate(CA-M), Cetyltrimethyl Ammonium Bromide(CTAB-M) and Cetylpyridinium Chloride and Sodium Dodecyl Sulfate(CS-M) [1-5]. Modified bentonites were studied by using XRD、FT-IR、SEM and EDS techniques to analysis their microstructure and surface morphology. The sorption of Se(IV) on inorganoand organo-bentonites were studied by batch experiment. The results showedinorgano-bentonites had higher sorption capacity than organo-bentonites. FeOH and La modified bentonites (FeOH-M and La-M) had highly removal efficiencyfor Se(IV) with 98% sorption.The sorption of organo-bentonites for Se(IV)were below 10%. SEM analyze results showed that the surface of AlOH-M and FeOH-M were flocculence and Fe-M surface was platelike. The sorption capacities for Se(IV) were FeOH-M with 11.9 mg/g , AlOH-M with 11.5 mg/g and Fe-M was 9 mg/g, respectively. It could be explained that flocculence structure has larger specific surface area and increases the sorption of Se(IV). Moreover, the sorption of Se(IV) on Mg-M was decreased with the increasing of ionic strength,whereasit had little effect on the other 4 kinds of inorgano-bentonites. It can be interpreted that the sorption of Se(IV) on Mg-M was formed an out-sphere complexation and others were formed inner-sphere complexations. Fig. 1.Sorption of Se(IV) on ionorgano- and organo-bentonites 149
  • [1]. Marco-Brown, J.L., et al., Adsorption of picloram herbicide on iron oxide pillared montmorillonite. Applied Clay Science, 2012. 58(0): p. 25-33. [2]. Meleshyn, A. and B. Riebe, Thermal Stability of Organoclays: Effects of Duration and Atmosphere of Isothermal Heating on Iodide Sorption. Environmental Science & Technology, 2010. 44(24): p. 9311-9317. [3]. Thakre, D., et al., Magnesium incorporated bentonite clay for defluoridation of drinking water. Journal of Hazardous Materials, 2010. 180(1–3): p. 122-130. [4]. Chen, D., et al., Characterization of anion–cationic surfactants modified montmorillonite and its application for the removal of methyl orange. Chemical Engineering Journal, 2011. 171(3): p. 1150-1158. [5]. Nguyen-Thanh, D., K. Block, and T.J. Bandosz, Adsorption of hydrogen sulfide on montmorillonites modified with iron. Chemosphere, 2005. 59(3): p. 343-353. 150
  • Pyrohydrolysis of Fluorides from Thorium-based Molten Salt Reactor Dong Xiaoyua,b, Liu Yuxiab, Zhang Lanb,* a b University of Chinese Academy of Sciences Shanghai Institute of Applied Physics, Chinese Academy of Sciences * Corresponding author: zhanglan@sinap.ac.cn Molten salt reactor is a promising nuclear reactor for future power production where liqu id fluoride salts are used as moderator and coolant. In order to take advantage of abundant tho rium resources, we are developing our own Th-based Molten Salt Reactor(TMSR). The post-p rocessing of molten salts is of great challenge. FLUOREX[1] is a hybrid system that combines fluoride volatility and solvent extraction methods. Between fluorination process and PUREX process, there is a pyrohydrolysis process where the fluoride compounds from fluorination process are converted to the oxides. Pyrohydrolysis[2] was first introduced by Warf et al. in 1950s, it has been proved to be an acceptable procedure being routinely used for separating halides from refractory matrices. It's a fast, reliable and convenient method for decomposition of solid samples. A new type of device for pyrohydrolysis research on fluorides has been developed in our laboratory. Now, stable running of the device can be guaranteed after a series of tests and deb ugging. We got exciting results about the pyrohydrolysis reaction of ZrF 4 . We are planning to take more experiments on other fluoride salts, such as UF 4 , ThF 4 and main fission products.T he aim is to find out the differences of their prohydrolytic characteristics, and achieve the pur pose of separation and recovery finally. [1] Hiroaki Kobayashi, Osamu Amano. et al. (2005) FLUOREX reprocessing system for the thermal reactors cycle and future thermal/fast reactors (coexistence) cycle . Progress in Nuclear Energy. 47:380-388. [2] Warf, J.C., Cline, W.D., Tevebaugh, R.D. (1954) Pyrohydrolysis in determination of fluoride and other halides. Anal. Chem. 26:342-346. Fig. 1. Outline of FLUOREX flow sheet 151
  • Comparative Study on Sorption of Eu(III) to Two Kinds of Mica: Muscovite and Phlogopite 1 2 Duoqiang Pan1,2, Zheming Wang2, Wangsuo Wu1 Radiochemistry Laboratory, Lanzhou University, Lanzhou 730000, China Pacific Northwest National Laboratory, Richland, Washington 99352, United State For the long-term deep geological disposal of radioactive waste, detailed information about geochemical behavior of radioactive under environmental conditions is necessary. The mobility of Eu(III) which is usually taken as a chemical homologue to Ln(III) and An(III) in natural media is significantly controlled by the sorption behavior that was influenced by Eu(III) speciation, sorbent properties and coexisteing organic/inorganic ligands. Understanding the Eu(III) sorption speciation and mechanism at the solid-liquid interface are imperative to predict the future retention/migration behavior of Ln(III)/An(III) in environmental media, and to provide reference for safety assessment of nuclear waste repository [1-3]. To figure out the detailed sequestration mechanism of Eu(III) in environmental media, many spectroscopic technologies (FTIR, XPS, EXAFS, TRLIFS) have been extensively employed to identify Eu(III) species and local atomic structures at the solidliquid interface at molecular level. TRLIF spectroscopy provides valuable information for speciation of fluorescent metal ions and the hydration number (n H2O ) can be obtained from the observed fluorescence decay rates, as H 2 O molecules in inner coordination spheres work as effective quenchers and that a change in n H2O is manifested in their decay lifetimes [4-7]. In this work, the interaction of Eu(III) and two types of prevalent rock forming mica (muscovite and phlogopite) were studied by using combination of batch and time resolved laser induced fluorescence (TRLIF) spectroscopy techniques. Effect of pH, ionic strength, and particularly the presence of organic and inorganic ligands on Eu(III) sorption were investigated by batch experiments. The number of H 2 Oin the first coordination sphere and the speciation of Eu(III) surface complexes were determined with the aid of TRLIF spectroscopy. The results shows that sorption of Eu(III) on phlogopite is much higher than that on muscovite. Eu(III) sorption on muscovite is influenced by both pH and ionic strength. Outersphere complexes through ion exchange dominate at the lower pH while inner-sphere complexation is the predominant mechanism at higher pH. The presence of HA promotes Eu(III) sorption at low pH by forming ternary surface complexes, while inhibites sorption at high pH because of formation of soluble binary complexes and the repulsive force between both negative charged HA and solid surface. PO 4 3- obviously enhances Eu(III) sorption, while SO 4 2- has a negative effect because of the weak affinity of SO 4 2- with solid surface as compared with PO 4 3-. Comparing with muscovite, the sorption capacity of phlogopite is much higher than that of muscovite. Superior sorption performance of phlogopite probably derives from the contribution of Fe-OH site on the surface of phlogopite, which is possibly responsible for the weaker fluorescence signal of Eu(III) adsorbed on it due to Fe fluorescence quenching effect. For results on phlogopite, the effects of PO 4 3- and SO 4 2- are similar to those on muscovite and HA and FA make little difference at low pH range while inhibit Eu(III) sorption at high pH, and the suppr ession of FA is more obvious than that of HA. 152
  • 100 Sorption % 80 60 40 20 NaCl 0.01 mol/L NaH2PO4 0.01 mol/L NaHSO4 0.01 mol/L HA 50mg/L 0 2 4 6 8 10 12 pH Figure 1 Effect of coexisting complex ligands on Eu(III) sorption to muscovite. 100 Sorption % 80 60 40 20 Phlogopite CEu=1E-05 mol/L Muscovite CEu=1E-06 mol/L Phlogopite CEu=1E-06 mol/L 0 2 4 6 8 10 12 pH Figure 2 Comparative sorption of Eu(III) on muscovite and phlogopite. Acknowledgement Financial Supports from National Natural Science Foundation of China (J1210001) and China Scholarship Council are acknowledged. [1] Rabung T., et al., Geochim Cosmochim Acta, 2005, 69(23): 5393-5402. [2] Janot N., Benedetti M.F., and Reiller P.E., Environ. Sci. Technol., 2011, 45: 3224-3230. [3] Fan Q.H., et al., Environ. Sci. Technol., 2009, 43: 5776-5782. [4] Ishida K., et al., J. Colloid Interf. Sci., 2012, 374: 258-266. [5] Stumpf T., et al., Environ. Sci. Technol., 2001, 35(18): 3691-3694. [6] Mandaliev P., et al., Geochim. Cosmochim. Acta, 2011, 75(8): 2017-2029. [7] Horrocks Jr. W. D. and Sudnick D. R., J. Am. Chem. Soc., 1979, 101(2): 334-340. 153
  • Sorption of Np(V) onto Na-bentonite: Effect of equilibrium time, pH, ionic strength and temperature Ping Lia, Zhi Liua, Zhijun Guoa, Quanlin Shib, Wangsuo Wua a Radiochemistry Laboratory, Lanzhou University, Lanzhou 730000, Gansu, China b Northwest Institute of Nuclear Technology, Xi’an 710024, China It has been estimated that 237Np will be a major contributor to environmental radioactivity from the disposal of high- or intermediate-level radioactive waste [1]. Neptunium(V), which is dominant under oxidizing conditions, shows weak interactions with mineral surfaces and is therefore regarded as a rather mobile species [2]. Therefore, it is very important to study the sorption/migration behavior of neptunium in repositories of nuclear waste. The aim of the present study is to obtain the sorption and diffusion data of long-lived neptunium(V) in the back-filled materials and natural wallrock. The sorption of Np(V) onto Na-bentonite was studied as a function of equilibrium time, pH, ionic strength and temperature under anaerobic conditions using a batch technique. The results showed that the Np(V) sorption onto Na-bentonite is strongly depended on solution pH, sorbent dose, Np(V) concentration, and weakly influenced by ionic strength. From Fig. 1 we can see that the sorption ratio increases with increasing sorbent dose. The calculated distribution coefficient (K d ) is weakly dependent on the sorbent dose, which is consistent with the physicochemical properties of K d value, i.e., K d is independent of solid-to-liquid ratio at very low solid content. The dependence of Np(V) sorption on ionic strength at different pH is presented in Fig. 2. A large ionic strength would restrict the sorption of Np(V) onto Na-bentonite at pH 6.5 while the sorption is independent of ionic strength at pH 8.5, which indicates that the sorption maybe of an outer-sphere complexation type at pH 6.5 and of an inner-sphere complexation type at pH 8.5 [3]. 100 Bentonite, pH=6.5 Bentonite, pH=8.5 80 60 40 Sorption (%) Sorption (%) 60 bentonite, pH=6.5 bentonite, pH=8.5 20 40 20 0 0 0 10 0.00 20 0.05 0.10 0.15 0.20 0.25 0.30 I (NaCl) / mol/L m/V (g/L) Fig. 2. Influence of ionic strength on sorption of Np(V) onto Na-bentonite. C 0 (Np(V))=4×10-7 mol/L, m/V=10 g/L. Fig. 1. Sorption of Np(V) onto Na-bentonite as a function of sorbent dose. C 0 (Np(V))=4×10-7 mol/L, I(NaCl)=0.1 mol/L. References: [1] P. Thakur, G.P. Mulholland, Appl Radiat Isotopes. 70 (2012). [2] K. Schmeide, G. Bernhard, Appl Geochem. 25 (2010). [3] K. Nakata, T. Fukuda, S. Nagasaki, S. Tanaka and A. Suzuki. Czech, J. Phys. 49/S1 (1999). 154
  • Application and Evaluation of Radioisotope in Tracer Technique Wei Qi, Juanjuan Bi, Peng Liu, Jing Wang, Wangsuo Wu* Radiochemistry Laboratory, Lanzhou University, Lanzhou, Gansu, China, 730000 * Corresponding author: wuws@lzu.edu.cn. As well known, radionuclides have a wide range of application in nuclear power, medical imaging, diagnosis and treatment of tumor and isotope tracer technique and so on. Especial in recently, because of the the advantages of high sensitivity, credibility, and freedom from interference, isotope tracer technique is known as a unique approach to investigate the fate and behavior of nuclide migration, drug metabolism, and nanoparticles in vitro or in vivo. Up to now, the application of isotope tracer technique in drug metabolism, nanoparticles and other materials in biological toxicology and pharmacology mainly obtain compounds /or chelations through chemical or physical synthetic methods. Though the yield and stability of the labeled compounds /or chelations has been proven very high in vitro, as well known, the environment of living organism is very complicated and highly variable in vivo, therefore, the efficiency and stability of the labeled compounds /or chelations are still under consideration, and the presence of radionuclides whether and how to effect on the fate and behavior of compounds /or chelations and the normal physiological function of experimental model. Consequently, the questions mentioned above will be analyzed and discussed on the basis of the application of isotope tracer technique in investigating the fate or behavior of nanoparticles in vivo in the review. Fig.1 Synthesis of the 14C-labeled fullerene 1a [1] . Refenence: 1. Shigeru Yamago, Hidetoshia Tokuyama, Eiichi Nakamura, Koichi Kikuchi, Shiji Kananishi, Keisuke Sueki, Hiromichi Nakahara, Shuichi Enomoto and Fumitoshi Ambe(1995). In vivo biological behavior of a water-miscible fullerene: 14C labeling, absorption, distribution, excretion and acute toxicity. Chemistry & Biology, 2:385-389. 155
  • Extraction of U(VI) and Th(IV) from aqueous solution into ionic liquid or n-pentanol using methylimidazole derivatives as extractants Wenkui Li, Yinglin Shen*, Wangsuo Wu Radiochemistry Laboratory, Lanzhou University, Lanzhou, Gansu, China, 730000 * Corresponding author : shenyl@lzu.edu.cn. As well known, the separation of uranium (U) and thorium (Th) has special significance due to long-term sustainability of the fuel cycle. Various analytical techniques are developed for it. In this work, a systematic study of the solvent extraction of uranium(VI) and thorium(IV) was performed using 1-methylimidazole (1-MZ) or 2-methylimidazole (2-MZ) as the extractants in ionic liquid or n-pentanol. Thorium (IV) was found to be quantitatively extracted with 1-MZ or 2-MZ in ionic liquid or n-pentanol media, and most of uranium(VI) stayed in the aqueous phase. The optimum conditions for extraction of these metals have been established by studying various parameters such as acid concentration, pH, reagent concentration, diluents, temperature and equilibration time. Slope analysis and ESR demonstrated that the extracted species are UO 2 -(MZ) (NO 3 ) 2 and Th-(MZ) 4 (NO 3 ) 4 , respectively in the organic phase. The extracted complexes were further charactered by UV and IR. It was observed that the separation factor (D Th /D U ) was 206 for thorium and uranium in ionic liquid. Similar results were achieved in n-pentanol using 1-MZ or 2-MZ as the extractants. These results indicate that MZ in ionic liquid or n-pentanol solvent system could be a potential candidate for separation of thorium and uranium. Fig. 1 Distribution ratios of U(VI) and Th(IV) with different concentrations of MZ extractant in ionic liquid media. 156